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Sample records for strong coolant depletion

  1. Depleted zinc oxide sertificate for safe zinc injection into water coolant of nuclear reactors

    International Nuclear Information System (INIS)

    Yurmanov, E.V.; Yurmanov, V.A.; Evropin, S.V.; Shut'ko, K.I.; Filimonov, S.V.; Zyryanov, S.M.; Timofeev, D.V.

    2015-01-01

    The prospects of the introduction of dosing of zinc depleted in the isotope 64Zn into the water coolant of WWER and RBMK are shown with the aim of removing the long-lived radionuclides accumulated in oxide films, including 60Co, which dominates in the radiation situation during parking for repairs and reloading of fuel, on the decommissioning of power units. The developed technical requirements for commercial zinc oxide powder include the allowable ranges of its granulometric characteristics (average and maximum pellet sizes) and isotopic composition (restriction of the share of the isotope 64Zn), its specific surface area and moisture content, as well as restrictions on the content of impurities to ensure its safe use in domestic NPP. The requirements for the purity of zinc oxide powder in the specification are developed taking into account the experience of its production at the enterprises of the State Corporation Rosatom for export deliveries to foreign nuclear power plants [ru

  2. Stagnation and Storage of Strongly Depleted Melts in Slow-Ultraslow Spreading Oceans: Evidence from the Ligurian Tethys

    Science.gov (United States)

    Piccardo, Giovanni; Guarnieri, Luisa; Padovano, Matteo

    2013-04-01

    Our studies of Alpine-Apennine ophiolite massifs (i.e., Lanzo, Voltri, Ligurides, Corsica) show that the Jurassic Ligurian Tethys oceanic basin was a slow-ultraslow spreading basin, characterized by the exposures on the seafloor of mantle peridotites with extreme compositional variability. The large majority of these peridotites are made of depleted spinel harzburgites and plagioclase peridotites. The former are interpreted as reactive peridotites formed by the reactive percolation of under-saturated, strongly trace element depleted asthenospheric melts migrated by porous flow through the mantle lithosphere. The latter are considered as refertilized peridotites formed by peridotite impregnation by percolated silica-saturated, strongly trace element depleted melts. Strongly depleted melts were produced as low-degrees, single melt increments by near fractional melting of the passively upwelling asthenosphere during the rifting stage of the basin. They escaped single melt increment aggregation, migrated isolated through the mantle lithosphere by reactive porous or channeled flow before oceanic opening, and were transformed into silica-saturated derivative liquids that underwent entrapment and stagnation in the shallow mantle lithosphere forming plagioclase-enriched peridotites. Widespread small bodies of strongly depleted gabbro-norites testify for the local coalescence of these derivative liquids. These melts never reached the surface (i.e., the hidden magmatism), since lavas with their composition have never been found in the basin. Subsequently, aggregated MORB melts upwelled within replacive dunite channels (as evidenced by composition of magmatic clinopyroxenes in dunites), intruded at shallow levels as olivine gabbro bodies and extruded as basaltic lavas, to form the crustal rocks of the oceanic lithosphere (i.e., the oceanic magmatism). Km-scale bodies of MORB olivine gabbros were intruded into the plagioclase-enriched peridotites, which were formed in the

  3. Mycolactone-Dependent Depletion of Endothelial Cell Thrombomodulin Is Strongly Associated with Fibrin Deposition in Buruli Ulcer Lesions.

    Directory of Open Access Journals (Sweden)

    Joy Ogbechi

    2015-07-01

    Full Text Available A well-known histopathological feature of diseased skin in Buruli ulcer (BU is coagulative necrosis caused by the Mycobacterium ulcerans macrolide exotoxin mycolactone. Since the underlying mechanism is not known, we have investigated the effect of mycolactone on endothelial cells, focussing on the expression of surface anticoagulant molecules involved in the protein C anticoagulant pathway. Congenital deficiencies in this natural anticoagulant pathway are known to induce thrombotic complications such as purpura fulimans and spontaneous necrosis. Mycolactone profoundly decreased thrombomodulin (TM expression on the surface of human dermal microvascular endothelial cells (HDMVEC at doses as low as 2 ng/ml and as early as 8 hrs after exposure. TM activates protein C by altering thrombin's substrate specificity, and exposure of HDMVEC to mycolactone for 24 hours resulted in an almost complete loss of the cells' ability to produce activated protein C. Loss of TM was shown to be due to a previously described mechanism involving mycolactone-dependent blockade of Sec61 translocation that results in proteasome-dependent degradation of newly synthesised ER-transiting proteins. Indeed, depletion from cells determined by live-cell imaging of cells stably expressing a recombinant TM-GFP fusion protein occurred at the known turnover rate. In order to determine the relevance of these findings to BU disease, immunohistochemistry of punch biopsies from 40 BU lesions (31 ulcers, nine plaques was performed. TM abundance was profoundly reduced in the subcutis of 78% of biopsies. Furthermore, it was confirmed that fibrin deposition is a common feature of BU lesions, particularly in the necrotic areas. These findings indicate that there is decreased ability to control thrombin generation in BU skin. Mycolactone's effects on normal endothelial cell function, including its ability to activate the protein C anticoagulant pathway are strongly associated with this

  4. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    military vehicles. Newer vehicles come factory- filled with ELC, while the Army continues to use traditional supplemental coolant additives (SCA)-based...UNCLASSIFIED TABLE OF CONTENTS EXTENDED LIFE COOLANT TESTING INTERIM REPORT TFLRF No. 478 by Gregory A. T. Hansen Edwin A...longer needed. Do not return it to the originator. UNCLASSIFIED UNCLASSIFIED EXTENDED LIFE COOLANT TESTING INTERIM REPORT TFLRF No

  5. Organic coolant summary report

    International Nuclear Information System (INIS)

    Smee, J.L.; Puttagunta, V.R.; Robertson, R.F.S.; Hatcher, S.R.

    1975-08-01

    This report summarizes the development of the use of organic liquids, specifically the terphenyls, as heat transfer mediums in nuclear reactors. All phases of the development program are covered, including the choice of coolant, decomposition, coolant reprocessing, physical properties, fouling, heat transfer, and corrosion. Few experimental results are given, but rather the emphasis is on summarizing the state of the art as it exists within AECL at the end of 1973. (Author)

  6. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  7. Environmentally Friendly Coolant System

    Energy Technology Data Exchange (ETDEWEB)

    David Jackson Principal Investigator

    2011-11-08

    Energy reduction through the use of the EFCS is most improved by increasing machining productivity. Throughout testing, nearly all machining operations demonstrated less land wear on the tooling when using the EFCS which results in increased tool life. These increases in tool life advance into increased productivity. Increasing productivity reduces cycle times and therefore reduces energy consumption. The average energy savings by using the EFCS in these machining operations with these materials is 9%. The advantage for end milling stays with flood coolant by about 6.6% due to its use of a low pressure pump. Face milling and drilling are both about 17.5% less energy consumption with the EFCS than flood coolant. One additional result of using the EFCS is improved surface finish. Certain machining operations using the EFCS result in a smoother surface finish. Applications where finishing operations are required will be able to take advantage of the improved finish by reducing the time or possibly eliminating completely one or more finishing steps and thereby reduce their energy consumption. Some machining operations on specific materials do not show advantages for the EFCS when compared to flood coolants. More information about these processes will be presented later in the report. A key point to remember though, is that even with equivalent results, the EFCS is replacing petroleum based coolants whose production produces GHG emissions and create unsafe work environments.

  8. Depleted uranium

    International Nuclear Information System (INIS)

    Huffer, E.; Nifenecker, H.

    2001-02-01

    This document deals with the physical, chemical and radiological properties of the depleted uranium. What is the depleted uranium? Why do the military use depleted uranium and what are the risk for the health? (A.L.B.)

  9. Reactor coolant pump flywheel

    Science.gov (United States)

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  10. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  11. Transient behaviour of main coolant pump in nuclear power plants

    International Nuclear Information System (INIS)

    Delja, A.

    1986-01-01

    A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

  12. Sodium as a reactor coolant

    International Nuclear Information System (INIS)

    Cesar, S.B.G.

    1989-01-01

    This work is related to the use of sodium as a reactor coolant, to the advantages and problems related to its use, its mechanical, thermophysics, eletronical, magnetic and nuclear properties. It is mainly a bibliographic review, with the aim of gathering the necessary information to persons initiating in the study of sodium and also as reference source. (author) [pt

  13. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  14. Molten Fuel-Coolant Interactions induced by coolant injection into molten fuel

    International Nuclear Information System (INIS)

    Park, H.S.; Yamano, Norihiko; Maruyama, Yu; Moriyama, Kiyofumi; Yang, Y.; Sugimoto, Jun

    1999-01-01

    To investigate Molten Fuel-Coolant Interactions (MFCIs) in various contact geometries, an experimental program, called MUSE (MUlti-configurations in Steam Explosions), has been initiated under the ALPHA program at JAERI in Japan. The first series of MUSE test has been focused on the coolant injection (CI) and stratified modes of FCIs using water as coolant and molten thermite as molten fuel. The effects of water jet subcooling, jet dynamics, jet shape and system constraint on FCIs energetic in these modes were experimentally investigated by precisely measuring their mechanical energy release in the MUSE facility. It was observed that measured mechanical energy increased with increasing of jet subcooling in a weakly constraint system but decreased in a strongly constraint system. FCI energetic also increased with increasing of water jet velocity. These results suggested that the penetration and dispersion phenomena of a water jet inside a melt determined the mixing conditions of FCIs in these contact modes and consequently played important roles on FCI energetics. To understand fundamental physics of these phenomena and possible mixing conditions in the MUSE tests, a set of visualization tests with several pairs of jet-pool liquids in non-boiling and isothermal conditions were carried out. Numerical simulations of a water jet penetrating into a water pool at non-boiling conditions showed similar behaviors to those observed in the visualization tests. (author)

  15. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  16. Coolant clean-up and recycle systems

    International Nuclear Information System (INIS)

    Ito, Takao.

    1979-01-01

    Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

  17. SMART core power control method by coolant temperature variation

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Cho, Byung Oh

    2001-08-01

    SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

  18. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  19. Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

    Science.gov (United States)

    Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

    2004-01-01

    The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

  20. Organic coolant for ARIES-III

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

    1991-04-01

    ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

  1. Physical properties of organic coolants

    International Nuclear Information System (INIS)

    Debbage, A.G.; Garton, D.A.; Kinneir, J.H.

    1963-03-01

    Density, viscosity, specific heat, vapour pressure and calorific value were measured within the temperature range 100 - 400 deg C for mixtures of Santowax R with pyrolytic high boiler and Santowax R with O.M.R.E. radiolytic high boiler; in addition measurements were made on Santowax OM, X-7 standard, X-7 loop coolant and O.M.R.E. coolant supplied by Atomic Energy of Canada Ltd. The accuracy of the measurements made were density (± 1/4%), viscosity (± 2%), specific heat (± 2%), vapour pressure (± 2%) and calorific value (± 1/2%). Thermal conductivity was calculated from an improved form of the Smiths equation with an accuracy within ± 6%. Equations fitted to the vapour pressure results were used to provide data outside the experimental range for burnout correlation purposes. The general effect of high boiler content on the specific heat and calorific values was small. The differences in physical property values for corresponding values of either pyrolytic or radiolytic high boiler were small for density (0.3%) and specific heat (2%), but quite large for viscosity (70%) with the pyrolytic high boiler mixture giving the higher value. The chemical analysis of all materials was based on gas chromatography and the relationship between this and an earlier distillation method established. (author)

  2. Cleaning of aluminum after machining with coolants

    International Nuclear Information System (INIS)

    Roop, B.

    1992-01-01

    An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

  3. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  4. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse

    2012-01-01

    The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat...... heat sink configurations reduces the coolant pumping power in the system....

  5. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

    Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

    1991-01-01

    In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

  6. Method of cleaning up coolants

    International Nuclear Information System (INIS)

    Asakura, Yamato; Kikuchi, Makoto; Uchida, Shunsuke.

    1979-01-01

    Purpose: To decrease the dose rates of the pipeways outside of reactors by charging adsorbents in a primary coolant circuits and recovering adsorbed solid impurities and ionic impurities in a reactor clean up system. Method: An addition device incorporating cleaning agents (adsorbents: metal oxide such as TiO 2 ) is connected to a feedwater pipe and a mixer for reacting the cleaning agent and solid impurities in feedwater is also provided. The cleaning agents added from the adding device into the feedwater are agitated in the mixer, adsorb the solid impurities in the feedwater and enter the reactor as reactor water while suspended in the feedwater. The reactor water is passed through reactor water clean up pipeway by a pump and removed with the cleaning agents and the solid impurities adsorbed thereon through the filter. (Seki, T.)

  7. Mathematical model of the reactor coolant pump

    International Nuclear Information System (INIS)

    Kozuh, M.

    1989-01-01

    The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

  8. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

  9. Flow boiling test of GDP replacement coolants

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.H. [comp.

    1995-08-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C{sub 4}F{sub 10} and C{sub 4}F{sub 8}, were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C{sub 4}F{sub 10} mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C{sub 4}F{sub 10} weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd.

  10. Continuous surveillance of reactor coolant circuit integrity

    International Nuclear Information System (INIS)

    1986-01-01

    Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

  11. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  12. Condition monitoring of main coolant pumps, Dhruva

    International Nuclear Information System (INIS)

    Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

    2002-01-01

    Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

  13. Coolant clean-up apparatus for reactors

    International Nuclear Information System (INIS)

    Yonekawa, Yasumasa; Akita, Minoru; Moritani, Kenji.

    1984-01-01

    Purpose: To prevent the occurrence of heat fatigue crackings due to the temperature difference between cleaned water and feed water at the portion on the pipeway where the two fluids are joined. Constitution: In the coolant clean-up system of a BWR type reactor, coolants are taken out from the side of the recycling system pump cooled by way of a regenerative heat exchanger and a non-regenerative heat exchanger and then flow into a filtering desalter for purification. Then, the purified water is joined with the feed water and returned to the inside of the reactor. Coolants for the auxiliary equipment cooler are recycled through non-regenerative heat exchanger and the flow control valve for the coolants is adjusted such that the temperature for the cleaned water accords with that of the feed water, whereby the temperature of the purified water is controlled to eliminate the temperature difference with the feed water. (Kamimura, M.)

  14. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    The invention deals with disengaging the coupling of a reactor coolant pump of a nuclear reactor feeding pressurized coolant. The disengaging coupling has two parts joined by bolts, at least one of them containing a driving agent within a bore. This is provided with a speed-depending ignition device in such manner that, if the critical speed is reached, the driving charge is ignited and the coupling is disengaged by destroying the bolts. (UWI) [de

  15. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  16. The operating reliability of the reactor coolant pump

    International Nuclear Information System (INIS)

    Grancy, W.

    1996-01-01

    There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

  17. Water coolant supply in relation to different ultrasonic scaler systems, tips and coolant settings

    NARCIS (Netherlands)

    Koster, T.J.G.; Timmerman, M.F.; Feilzer, A.J.; van der Velden, U.; van der Weijden, F.A.

    2009-01-01

    Objective: This study evaluated "in vitro" the consistency of the water coolant supply for five ultrasonic scaler systems in relation to the tip type and different coolant settings. Material and Methods: The systems were: EMS PM-400, EMS PM-600, Satelec P-max, Dürr Vector and Dentsply Cavitron. For

  18. NGNP Reactor Coolant Chemistry Control Study

    International Nuclear Information System (INIS)

    Castle, Brian

    2010-01-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor. The 'NGNP Reactor Coolant Chemistry Control Study' has identified the necessary components and the approach used to control the coolant chemistry in high-temperature gas-cooled reactors (HTGR) for eventual use in the Next Generation Nuclear Plant (NGNP). The focus of this study is to determine what technical developments have taken place since previous helium purification systems (HPSs) were operated in previous HTGRs. Since previous HTGR operational experience, a considerable number of corrosion studies have been performed on various high temperature materials in helium environments with various chemistries at elevated temperatures. This newer work has been reviewed and incorporated into this corrosion study to determine how this new information impacts the design requirements for modern HPSs. Research shows that there are ratios of key impurities that determine whether or not a protective oxide layer is formed or if a deleterious relationship is formed between the material and the coolant. These and other influencing factors are taken into consideration when considering the design requirements for the HPS for the NGNP. In addition to the metallic components in the primary circuit, there are other important material considerations that greatly impact the chemistry requirements for the helium coolant, which are the graphite core support structures. While the metallic components need a small amount of moisture in the coolant to form a protective oxide layer, the graphite needs a very dry environment to prevent degradation. Recommendations and conclusion are made that identify important HPS parameters that should be considered in the design of the HPS for the NGNP. Following the coolant chemistry

  19. Lead Coolant Test Facility Development Workshop

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz

    2005-06-01

    A workshop was held at the Idaho National Laboratory on May 25, 2005, to discuss the development of a next generation lead or lead-alloy coolant test facility. Attendees included representatives from the Generation IV lead-cooled fast reactor (LFR) program, Advanced Fuel Cycle Initiative, and several universities. Several participants gave presentations on coolant technology, existing experimental facilities for lead and lead-alloy research, the current LFR design concept, and a design by Argonne National Laboratory for an integral heavy liquid metal test facility. Discussions were focused on the critical research and development requirements for deployment of an LFR demonstration test reactor, the experimental scope of the proposed coolant test facility, a review of the Argonne National Laboratory test facility design, and a brief assessment of the necessary path forward and schedule for the initial stages of this development project. This report provides a summary of the presentations and roundtable discussions.

  20. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  1. Main coolant pump testing at Ontario Hydro

    International Nuclear Information System (INIS)

    Hartlen, R.

    1991-01-01

    This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

  2. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  3. The upgrade of intense pulsed neutron source (IPNS) through the change of coolant and reflector

    CERN Document Server

    Baek, I C; Iverson, E B

    2002-01-01

    The current intense pulsed neutron source (IPNS) depleted uranium target is cooled by light water. The inner reflector material is graphite and the outer reflector material is beryllium. The presence of H sub 2 O in the target moderates neutrons and leads to a higher absorption loss in the target than is necessary. D sub 2 O coolant in the small quantities required minimizes this effect. We have studied the possible improvement in IPNS beam fluxes that would result from changing the coolant from H sub 2 O to D sub 2 O and the inner reflector from graphite to beryllium. Neutron intensities were calculated for directions normal to the viewed surface of each moderator for four different cases of combinations of target coolant and reflector materials. The simulations reported here were performed using the MCNPX (version 2.1.5) computer program. Our results show that substantial gains in neutron beam intensities can be achieved by appropriate combination of target coolant and reflector materials. The combination o...

  4. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  5. Fission product release into the primary coolant

    International Nuclear Information System (INIS)

    Apperson, C.E.

    1977-01-01

    The analytic evaluation of steady state primary coolant activity is discussed. The reported calculations account for temperature dependent fuel failure in two particle types and arbitrary radioactive decay chains. A matrix operator technique implemented in the SUVIUS code is used to solve the simultaneous equations. Results are compared with General Atomic Company's published results

  6. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  7. Internal twist drill coolant channel modelling using computational fluid dynamics

    OpenAIRE

    Johns, AS; Hewson, RW; Merson, E; Summers, JL; Thompson, HM

    2014-01-01

    Coolant technologies have become increasingly used within twist-drill machining. Internal helical coolant channels are a common means of delivering high-pressure coolant directly to the cutting edge to increase the transfer of heat away from the cutting edge and to aid with chip evacuation. However, the complex behaviour of Coriolis and centrifugal forces generated by large angular velocities and the curvature of the coiled channel govern the highly turbulent flow of coolant. This research us...

  8. Experimental and numerical investigation of the coolant mixing during fast deboration transients

    International Nuclear Information System (INIS)

    Hoehne, T.; Rohde, U.; Weiss, F.P.

    1999-01-01

    For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the VVER-440 were performed with a CFD code (CFX-4). The comparison with experimental data and an analytical mixing model which is implemented in the neutron-kinetic code DYN3D showed a good agreement for near-nominal conditions. First experiments at the Rossendorf Mixing Test Facility ROCOM were performed simulating the start-up of the first main coolant pump. The reference reactor for the geometrically 1:5 scaled Plexiglas model is the German Konvoi type PWR. After demonstrating the capability of the CFD code to simulate these complicated flow transients, calculations were performed for the start-up of the first pump in a VVER-440 type reactor. These calculations are a first step of understanding the coolant mixing in the RPV of a VVER-440 type reactor under transient conditions. The results of the calculation show a very complex flow in the downcomer. A high downcomer of VVER-440 and the existence of the lower control rod chamber support coolant mixing is concluded. (author)

  9. Steady MHD flow for molten metal coolants in advanced nuclear systems

    International Nuclear Information System (INIS)

    Koreshi, Z.U.

    2004-01-01

    The flow of a molten metal coolant, incorporating viscosity and a magnetic force in a channel typical of a nuclear fusion reactor is formulated. This is based on the flow conservation equations (continuity and momentum), along with Maxwell's equations and Ohm's law. These equations are solved with assumptions of steady fully-developed flow for Newtonian fluids. Exact solutions are compared with a Direct Numerical Simulation (DNS) to compute important engineering design parameters such as velocity distribution and pressure drop in a coolant channel. The analysis can be used for several candidate coolants including, molten lithium, sodium, lead, and compounds. The difference in velocity profile, and hence pressure drop, is strongly dependent on the magnetic field. In fact, it is shown that a flat (almost-turbulent) profile is obtained for the flow. (author)

  10. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    Science.gov (United States)

    Rezania, A.; Rosendahl, L. A.

    2012-06-01

    The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The three-dimensional governing equations for the fluid flow and the heat transfer are solved using the finite-volume method for a wide range of pressure drop laminar flows along the heat sink. The temperature and the mass flow rate distribution in the heat sink are discussed. The results, which are in good agreement with previous computational studies, show that using suggested heat sink configurations reduces the coolant pumping power in the system.

  11. The 10B(n,α)7Li reaction in PWR coolants: calculations of the effect on coolant pH and on decreases in 10B isotopic fractions

    International Nuclear Information System (INIS)

    Polley, M.V.

    1988-07-01

    Boron is used as a chemical shim in PWRs for reactivity control and is added in the form of boric acid to the primary coolant. The 10 B(n,α) 7 Li reaction leads to a continuous increase in 7 Li in the primary coolant and to a continuous decrease in 10 B the isotope of boron responsible for control of reactivity. The rate of increase in coolant pH due to 7 Li production is calculated for the Sizewell 'B' PWR to enable judgements to be made on the frequency of sampling and removal of lithium required to maintain the pH of the primary coolant within the desired limits. Calculations are contrasted for the cases of natural boron and 100% 10 B chemical shims, for both a normal cycle and an extended 18 month cycle. Calculations of 10 B depletion over 30 years of operation as a function of the quantity of boron discharged to waste are also presented. 10 B isotopic fractions are calculated for the reactor coolant (RC), boric acid tanks (BATs) and refuelling water storage tank (RWST) assuming rapid mixing of BAT and RC boron for tritium control and other reasons. Such predictions enable assessments of the reactor physics implications of 10 B consumption to be made. (author)

  12. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  13. Recovery studies for plutonium machining oil coolant

    International Nuclear Information System (INIS)

    Navratil, J.D.; Baldwin, C.E.

    1977-01-01

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products

  14. Design of automotive engine coolant hoses

    Directory of Open Access Journals (Sweden)

    Hrishikesh D BACHCHHAV

    2018-03-01

    Full Text Available In this paper, we are present the performance of engine coolant hoses (radiator hoses used in passenger cars by checking various physical behaviours such as hose leakage, hose burst, hose collapse or any mechanical damage as studied-thru design guidelines, CFD analysis and product validation testing and also check pressure drop of the hoses when engine will be running. The design term is more likely used for technical part modelling using CAD tool. Later on, we will focus on the transformation of the part design to process design. The process design term is more likely used for "tooling design" for manufacturing of the product using CAD Tool. Then inlet hose carries coolant from engine to radiator inlet tank, then coolant circulated in radiator and passed through radiator outlet tank to water pump of engine with the help of outlet hose. After that …nding any leakage, Burst, damage or collapse of hose and pressure drop of the hose with the help of design checklist, CFD Analysis and product validation testing.

  15. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  16. Pre-transplant donor-specific T-cell alloreactivity is strongly associated with early acute cellular rejection in kidney transplant recipients not receiving T-cell depleting induction therapy.

    Directory of Open Access Journals (Sweden)

    Elena Crespo

    Full Text Available Preformed T-cell immune-sensitization should most likely impact allograft outcome during the initial period after kidney transplantation, since donor-specific memory T-cells may rapidly recognize alloantigens and activate the effector immune response, which leads to allograft rejection. However, the precise time-frame in which acute rejection is fundamentally triggered by preformed donor-specific memory T cells rather than by de novo activated naïve T cells is still to be established. Here, preformed donor-specific alloreactive T-cell responses were evaluated using the IFN-γ ELISPOT assay in a large consecutive cohort of kidney transplant patients (n = 90, to assess the main clinical variables associated with cellular sensitization and its predominant time-frame impact on allograft outcome, and was further validated in an independent new set of kidney transplant recipients (n = 67. We found that most highly T-cell sensitized patients were elderly patients with particularly poor HLA class-I matching, without any clinically recognizable sensitizing events. While one-year incidence of all types of biopsy-proven acute rejection did not differ between T-cell alloreactive and non-alloreactive patients, Receiver Operating Characteristic curve analysis indicated the first two months after transplantation as the highest risk time period for acute cellular rejection associated with baseline T-cell sensitization. This effect was particularly evident in young and highly alloreactive individuals that did not receive T-cell depletion immunosuppression. Multivariate analysis confirmed preformed T-cell sensitization as an independent predictor of early acute cellular rejection. In summary, monitoring anti-donor T-cell sensitization before transplantation may help to identify patients at increased risk of acute cellular rejection, particularly in the early phases after kidney transplantation, and thus guide decision-making regarding the use of induction

  17. Enriched boric acid as an optimized neutron absorber in the EPR primary coolant

    International Nuclear Information System (INIS)

    Cosse, Christelle; Jolivel, Fabienne; Berger, Martial

    2012-09-01

    reasons, the EBA Boron 10 isotopic atomic abundance target value is 37 at.%. However, EBA conditioning also necessitates the renewal of boric acid supply, operating and management schemes. Supplies must cover the needs for initial conditioning and for later compensation of losses during operation. These losses can be divided into three categories: 1. volumetric losses (such as leaks) 2. decrease of total Boron concentration (dilution for example) 3. Boron neutron core depletion (through Boron 10 consuming reactions). The first two losses are classically compensated with EBA supplied with the required isotopic abundance. Very Enriched boric acid (VEBA) will also be supplied to compensate Boron 10 enrichment losses due to neutron depletion ( 10 B [n, He] 7 Li). According to an optimized boron neutron depletion management scheme, neutron depletion will be directly compensated into boric acid make-up tanks prior to any refuelling shutdown, by injection of reasonable quantities of VEBA (Boron 10 atomic abundance superior to 90 at.%). Finally, EPR operators and chemists will have to cope with these two different sorts of boric acid. Distinctive methods will be employed to avoid any error: use of an optimized and strict policy of different units of Boron used in instrumentation and control systems and in operating procedures; distinct packaging and storage; separate procedures; specific skills for Boron neutron core depletion management. EPR operation has to rely on appropriate parameters and procedures. Total Boron concentration and Boron 10 isotopic abundance monitoring will be periodically achieved by sampling and analysing all borated systems, so called 'boron mapping'. The Boron 10 concentration is continuously monitored in the reactor coolant system with the on-line neutronic boron meter located in the Nuclear Sampling System (NSS): this data is directly used for reactor operation and chemistry control. The use of EBA in the EPR is one of the major progresses for

  18. Coolant Void Reactivity Analysis of CANDU Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

  19. Multirods burst tests under loss-of-coolant conditions

    International Nuclear Information System (INIS)

    Kawasaki, S.; Uetsuka, H.; Furuta, T.

    1983-01-01

    In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

  20. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  1. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  2. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  3. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  4. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  5. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

    The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

  6. Addressing Ozone Layer Depletion

    Science.gov (United States)

    Access information on EPA's efforts to address ozone layer depletion through regulations, collaborations with stakeholders, international treaties, partnerships with the private sector, and enforcement actions under Title VI of the Clean Air Act.

  7. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  8. Slow coolant phaseout could worsen warming

    Science.gov (United States)

    Reese, April

    2018-03-01

    In the summer of 2016, temperatures in Phalodi, an old caravan town on a dry plain in northwestern India, reached a blistering 51°C—a record high during a heat wave that claimed more than 1600 lives across the country. Wider access to air conditioning (AC) could have prevented many deaths—but only 8% of India's 249 million households have AC. As the nation's economy booms, that figure could rise to 50% by 2050. And that presents a dilemma: As India expands access to a life-saving technology, it must comply with international mandates—the most recent imposed just last fall—to eliminate coolants that harm stratospheric ozone or warm the atmosphere.

  9. Power module assemblies with staggered coolant channels

    Science.gov (United States)

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  10. Coolant sampling device for a reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Ko.

    1982-01-01

    Purpose: To enable accurate measurement of fine quantity of metal impurity in primary coolant by removing unneccessary impurity adhered to a collecting tube when the metal impurity is detected. Constitution: A sampling device is mounted at a recirculating system pipe in a BWR type reactor, a collecting tube is connected to the pipe, a sampled water collecting nozzle is also mounted, and a sluice valve for flowing and shutting off the sampled water is disposed in the vicinity of the nozzle. Acid cleaning liquid which is bubbled by inert gas from a tube provided at the downstream side of the valve cleans the inner surface of the collecting tube over the entire length, thereby effectively removing impurity adhered onto the inner surface of the tube. (Kamimura, M.)

  11. Automatic coolant flow control device for a nuclear reactor assembly

    Science.gov (United States)

    Hutter, E.

    1984-01-27

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  12. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  13. Causes and effects of coolant parameters oscillations on NPP with light water reactor

    International Nuclear Information System (INIS)

    Proskuriakov, K.N.

    2004-01-01

    Paper describes some original academic and engineering R and D results carried out in NPPD MPEI (TU) which that time advanced the current requirements of nuclear power. Acoustic method of early detection of steam cavity appearance in WWER and PWR reactor core had been published before the accident at TMI-2. Specialists recognized this method capability for prevention of one of the most severe accident. Subject of our researches is clearing up of the nature and frequencies of coolant parameters oscillations in operational and emergency situations. Calculated eigenfrequencies of self-excited coolant pressure oscillations based on carry out models had been verified by measurements on NPP with WWER-1000. Spectrum identification (most possible in presence) is provided. The dependence of coolant pressure oscillation amplitudes from reactor power level was revealed. As the probable mechanism of dynamic stresses occurrence in the hot collector's welded line SG-4, the parametrical resonance concept is offered. Bring forward CSGDPS that in particular shows the strong influence of hydraulic resistances on two phases flow stability. As CSGDPS much more than in real NPP systems (due to difference in hydraulic resistances) it is reasonable to realize of PSIS experimental units' reconstruction. The objective of our initiative is to build the computer code for calculating the coolant pressure eigenfrequencies oscillations at normal and emergency operation. This code could provide possibility of NPP vibroacoustical certification. Also the objective of R and D is theoretically substantiation and determination the control actions for resonance prevention between coolant and structures (fuel assemblies, reactor internals and equipment), the service life prolongation, the improvement of NPP reliability and the safety. (author)

  14. Modeling the spatial distribution of the parameters of the coolant in the reactor volume

    International Nuclear Information System (INIS)

    Nikonov, S.P.

    2011-01-01

    In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (Author)

  15. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  16. Method of charging instruments into liquid metal coolant

    International Nuclear Information System (INIS)

    Yamazaki, Hiroshi

    1980-01-01

    Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

  17. Determination of iron content in the nitrogen tetroxide coolant

    International Nuclear Information System (INIS)

    Kovaleva, E.P.; Bordak, L.V.

    1979-01-01

    When utilizing dissociating nitrogen tetroxide as a coolant and working fluid of NPP, it is necessary to work out the methods of control of the coolant quality characteristic. The important characteristic of the coolant quality on impurities is the content of corrosion products in it and iron in particular, as the basic macrocomponent of structural materials. For determination of iron content in the nitrogen tetroxide-based coolant, sulphosalicylic method in ammoniacal variant is proposed. A reflected the experimental working outs of the o-phenanthroline method and the method with sulphosalicylic acid. The method of the sample preparation of the N 2 O 4 -based coolant for the subsequent calorimetric analysis is cited

  18. Revisiting The Depleted Self.

    Science.gov (United States)

    Abraham, Reggie

    2018-04-01

    This article revisits Donald Capps's book The Depleted Self (The depleted self: sin in a narcissistic age. Fortress Press, Minneapolis, 1993), which grew out of his 1990 Schaff Lectures at Pittsburgh Theological Seminary. In these lectures Capps proposed that the theology of guilt had dominated much of post-Reformation discourse. But with the growing prevalence of the narcissistic personality in the late twentieth century, the theology of guilt no longer adequately expressed humanity's sense of "wrongness" before God. Late twentieth-century persons sense this disjunction between God and self through shame dynamics. Narcissists are not "full" of themselves, as popular perspectives might indicate. Instead, they are empty, depleted selves. Psychologists suggest this stems from lack of emotional stimulation and the absence of mirroring in the early stages of life. The narcissist's search for attention and affirmation takes craving, paranoid, manipulative, or phallic forms and is essentially a desperate attempt to fill the internal emptiness. Capps suggests that two narratives from the Gospels are helpful here: the story of the woman with the alabaster jar and the story of Jesus's dialogue with Mary and John at Calvary. These stories provide us with clues as to how depleted selves experienced mirroring and the potential for internal peace in community with Jesus.

  19. Ozone depletion update.

    Science.gov (United States)

    Coldiron, B M

    1996-03-01

    Stratospheric ozone depletion due to chlorofluorocarbons an d increased ultraviolet radiation penetration has long been predicted. To determine if predictions of ozone depletion are correct and, if so, the significance of this depletion. Review of the English literature regarding ozone depletion and solar ultraviolet radiation. The ozone layer is showing definite thinning. Recently, significantly increased ultraviolet radiation transmission has been detected at ground level at several metering stations. It appears that man-made aerosols (air pollution) block increased UVB transmission in urban areas. Recent satellite measurements of stratospheric fluorine levels more directly implicate chlorofluorocarbons as a major source of catalytic stratospheric chlorine, although natural sources may account for up to 40% of stratospheric chlorine. Stratospheric chlorine concentrations, and resultant increased ozone destruction, will be enhanced for at least the next 70 years. The potential for increased transmission of ultraviolet radiation will exist for the next several hundred years. While little damage due to increased ultraviolet radiation has occurred so far, the potential for long-term problems is great.

  20. Experimental and numerical investigation of the coolant mixing during fast deboration transients

    International Nuclear Information System (INIS)

    Hoehne, T.; Rohde, U.; Weiss, F.P.

    1999-01-01

    For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the WWER-440 were performed with a CFD code (CFX-4). For this calculation the RPV from the cold legs inlet through the downcomer, the lower plenum and the lower core support plate was nodulized in detail. The comparison with experimental data and analytical mixing model which is implemented in the neutron kinetic code DYN3D showed a good agreement for near-nominal conditions (all MCPs are running). The comparison between the CFD-results and the analytical model revealed differences for MSLB conditions[1]. (Authors)

  1. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  2. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

    2013-01-01

    Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

  3. SUBSTATIONS OF DISTRICT HEATING SYSTEMS WITH PULSE COOLANT CIRCULATION

    Directory of Open Access Journals (Sweden)

    Andrey N. Makeev

    2017-01-01

    Full Text Available Abstract. Objectives The aim of the study is to generalise the results of the application of technologies and means for organising pulse coolant flow within a district heating system in order to increase its energy efficiency based on the organisation of local hydraulic shocks and the subsequent use of their energy to ensure the purification of heat energy equipment, intensify the heat transfer process and realise the possibility of transforming the available head from one hydraulic circuit to another. Methods Substations connecting the thermal power installations of consumers with heat networks via dependent and independent schemes are analytically generalised. The use of pulse coolant circulation is proposed as a means of overcoming identified shortcomings. Results Principal schemes of substations with pulse coolant circulation for dependent and independent connection of thermal power installations are detailed. A detailed description of their operation is given. The advantages of using pulse coolant circulation in substations are shown. The materials reflecting the results of the technical implementation and practical introduction of this technology are presented. Conclusion Theoretical analysis of the operation of the basic schemes of substations with pulse coolant circulation and the results of their practical application, as well as the materials of scientific works devoted to the use of the energy of a hydraulic impact and the study of the effect of pulse coolant flow on thermal and hydrodynamic processes, have yielded a combination of factors reflecting technical and economic rationality of application of pulse coolant circulation. 

  4. Development of lead-bismuth coolant technology for nuclear device

    International Nuclear Information System (INIS)

    Kamata, Kin-ya; Kitano, Teruaki; Ono, Mikinori

    2004-01-01

    Liquid lead-bismuth is a promising material as a future fast reactor coolant or an intensive neutron source material for accelerator driven transmutation system (ADS). To develop nuclear plants and their installations using lead-bismuth coolant for practical use, both coolant technologies, inhabitation process of steels and quality control of coolant, and total operation system for liquid lead-bismuth plants are required. Based on the experience of liquid metal coolant, Mitsui Engineering and Shipbuilding Co., Ltd. (MES) has completed the liquid lead-bismuth forced circulation loop and has acquired various engineering data on main components including economizer. As a result of tis operation, MES has developed key technologies of lead-bismuth coolant such as controlling of oxygen content in lead-bismuth and a purification of lead-bismuth coolant. MES participated in the national project, ''The Development of Accelerator Driven Transmutation System'', together with JAERI (Japan Atomic Energy Research Institute) and started corrosion test for beam window of ADS. (author)

  5. Depletion sensitivity predicts unhealthy snack purchases.

    Science.gov (United States)

    Salmon, Stefanie J; Adriaanse, Marieke A; Fennis, Bob M; De Vet, Emely; De Ridder, Denise T D

    2016-01-01

    The aim of the present research is to examine the relation between depletion sensitivity - a novel construct referring to the speed or ease by which one's self-control resources are drained - and snack purchase behavior. In addition, interactions between depletion sensitivity and the goal to lose weight on snack purchase behavior were explored. Participants included in the study were instructed to report every snack they bought over the course of one week. The dependent variables were the number of healthy and unhealthy snacks purchased. The results of the present study demonstrate that depletion sensitivity predicts the amount of unhealthy (but not healthy) snacks bought. The more sensitive people are to depletion, the more unhealthy snacks they buy. Moreover, there was some tentative evidence that this relation is more pronounced for people with a weak as opposed to a strong goal to lose weight, suggesting that a strong goal to lose weight may function as a motivational buffer against self-control failures. All in all, these findings provide evidence for the external validity of depletion sensitivity and the relevance of this construct in the domain of eating behavior. Copyright © 2015 Elsevier Ltd. All rights reserved.

  6. MIC damage in a water coolant header for remote process equipment

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1994-01-01

    Stainless steel water piping used to supply coolant for remote chemical separations equipment developed leaks during low flow conditions resulting from an extended interruption of operations. All the leaks occurred at welds in the bottom zone of the pipe, which was blanketed with silt deposits from the unfiltered well water used for cooling. Ultrasonic, radiographic, and metallographic examinations of leak sites revealed worm hole pitting adjacent to the welds. Seepage at the penetrations was strongly acidic and resulted in corrosion on the external pipe surfaces beneath brown crusty deposits which had developed. Analyses of the water and deposits suggest a strong propensity toward microbiologically influenced corrosion (MIC) and fouling

  7. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1984-01-01

    A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications. (orig.)

  8. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1986-01-01

    A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications

  9. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

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

  10. Numerical model simulation of atmospheric coolant plumes

    International Nuclear Information System (INIS)

    Gaillard, P.

    1980-01-01

    The effect of humid atmospheric coolants on the atmosphere is simulated by means of a three-dimensional numerical model. The atmosphere is defined by its natural vertical profiles of horizontal velocity, temperature, pressure and relative humidity. Effluent discharge is characterised by its vertical velocity and the temperature of air satured with water vapour. The subject of investigation is the area in the vicinity of the point of discharge, with due allowance for the wake effect of the tower and buildings and, where application, wind veer with altitude. The model equations express the conservation relationships for mometum, energy, total mass and water mass, for an incompressible fluid behaving in accordance with the Boussinesq assumptions. Condensation is represented by a simple thermodynamic model, and turbulent fluxes are simulated by introduction of turbulent viscosity and diffusivity data based on in-situ and experimental water model measurements. The three-dimensional problem expressed in terms of the primitive variables (u, v, w, p) is governed by an elliptic equation system which is solved numerically by application of an explicit time-marching algorithm in order to predict the steady-flow velocity distribution, temperature, water vapour concentration and the liquid-water concentration defining the visible plume. Windstill conditions are simulated by a program processing the elliptic equations in an axisymmetrical revolution coordinate system. The calculated visible plumes are compared with plumes observed on site with a view to validate the models [fr

  11. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  12. Corrosion problems with aqueous coolants, final report

    Energy Technology Data Exchange (ETDEWEB)

    Diegle, R B; Beavers, J A; Clifford, J E

    1980-04-11

    The results of a one year program to characterize corrosion of solar collector alloys in aqueous heat-transfer media are summarized. The program involved a literature review and a laboratory investigation of corrosion in uninhibited solutions. It consisted of three separate tasks, as follows: review of the state-of-the-art of solar collector corrosion processes; study of corrosion in multimetallic systems; and determination of interaction between different waters and chemical antifreeze additives. Task 1 involved a comprehensive review of published literature concerning corrosion under solar collector operating conditions. The reivew also incorporated data from related technologies, specifically, from research performed on automotive cooling systems, cooling towers, and heat exchangers. Task 2 consisted of determining the corrosion behavior of candidate alloys of construction for solar collectors in different types of aqueous coolants containing various concentrations of corrosive ionic species. Task 3 involved measuring the degradation rates of glycol-based heat-transfer media, and also evaluating the effects of degradation on the corrosion behavior of metallic collector materials.

  13. Intrinsic Depletion or Not

    DEFF Research Database (Denmark)

    Klösgen, Beate; Bruun, Sara; Hansen, Søren

      The presence of a depletion layer of water along extended hydrophobic interfaces, and a possibly related formation of nanobubbles, is an ongoing discussion. The phenomenon was initially reported when we, years ago, chose thick films (~300-400Å) of polystyrene as cushions between a crystalline...... carrier and biomimetic membranes deposited thereupon and exposed to bulk water. While monitoring the sequential build-up of the sandwiched composite structure by continuous neutron reflectivity experiments the formation of an unexpected additional layer was detected (1). Located at the polystyrene surface...... in between he polymer cushion and bulk water the layer was attributed to water of reduced density and was called "depletion layer".  Impurities or preparative artefacts were excluded as its origin. Later on, the formation of nanobubbles from this vapour-like water phase was initiated by tipping the surface...

  14. Capital expenditure and depletion

    International Nuclear Information System (INIS)

    Rech, O.; Saniere, A.

    2003-01-01

    In the future, the increase in oil demand will be covered for the most part by non conventional oils, but conventional sources will continue to represent a preponderant share of the world oil supply. Their depletion represents a complex challenge involving technological, economic and political factors. At the same time, there is reason for concern about the decrease in exploration budgets at the major oil companies. (author)

  15. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

    2001-07-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

  16. Comparative analysis of coolants for FBR of future nuclear power

    International Nuclear Information System (INIS)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

    2001-01-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

  17. High converter pressurized water reactor with heavy water as a coolant

    International Nuclear Information System (INIS)

    Ronen, Y.; Reyev, D.

    1983-01-01

    There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

  18. Ozone-depleting Substances (ODS)

    Data.gov (United States)

    U.S. Environmental Protection Agency — This site includes all of the ozone-depleting substances (ODS) recognized by the Montreal Protocol. The data include ozone depletion potentials (ODP), global warming...

  19. Failure to Replicate Depletion of Self-Control

    Science.gov (United States)

    Xu, Xiaomeng; Demos, Kathryn E.; Leahey, Tricia M.; Hart, Chantelle N.; Trautvetter, Jennifer; Coward, Pamela; Middleton, Kathryn R.; Wing, Rena R.

    2014-01-01

    The limited resource or strength model of self-control posits that the use of self-regulatory resources leads to depletion and poorer performance on subsequent self-control tasks. We conducted four studies (two with community samples, two with young adult samples) utilizing a frequently used depletion procedure (crossing out letters protocol) and the two most frequently used dependent measures of self-control (handgrip perseverance and modified Stroop). In each study, participants completed a baseline self-control measure, a depletion or control task (randomized), and then the same measure of self-control a second time. There was no evidence for significant depletion effects in any of these four studies. The null results obtained in four attempts to replicate using strong methodological approaches may indicate that depletion has more limited effects than implied by prior publications. We encourage further efforts to replicate depletion (particularly among community samples) with full disclosure of positive and negative results. PMID:25333564

  20. Failure to replicate depletion of self-control.

    Directory of Open Access Journals (Sweden)

    Xiaomeng Xu

    Full Text Available The limited resource or strength model of self-control posits that the use of self-regulatory resources leads to depletion and poorer performance on subsequent self-control tasks. We conducted four studies (two with community samples, two with young adult samples utilizing a frequently used depletion procedure (crossing out letters protocol and the two most frequently used dependent measures of self-control (handgrip perseverance and modified Stroop. In each study, participants completed a baseline self-control measure, a depletion or control task (randomized, and then the same measure of self-control a second time. There was no evidence for significant depletion effects in any of these four studies. The null results obtained in four attempts to replicate using strong methodological approaches may indicate that depletion has more limited effects than implied by prior publications. We encourage further efforts to replicate depletion (particularly among community samples with full disclosure of positive and negative results.

  1. Experimental research to investigate the performance of bio coolant when turning of mild carbon steel

    Science.gov (United States)

    Agus Susanto, Tri; Nur, Rusdi

    2017-04-01

    Some literatures have been reported that the using bio coolant show better lubricating and cooling performances and reduce the occupational health risks associated with petroleum-oil-based coolant since they have lower toxicity. This paper investigates the effect the cutting conditions on the surface roughness through turning of mild carbon steel using dry, coolant and bio coolant. Measurement of surface roughness was conducted and then compared with the change of the cutting conditions. The relationship between surface roughness and cutting conditions was created in a curve for different of the cutting speed and coolant. The results indicate that the surface roughness was reduced when the speed of cutting is set to the highest level for all of coolant conditions (dry, coolant, and bio coolant) and constant of DOC and feed. The surface roughness had better performance using bio coolant than coolant conventional (mineral fluid).

  2. Consequences of biome depletion

    International Nuclear Information System (INIS)

    Salvucci, Emiliano

    2013-01-01

    The human microbiome is an integral part of the superorganism together with their host and they have co-evolved since the early days of the existence of the human species. The modification of the microbiome as a result changes in food and social habits of human beings throughout their life history has led to the emergence of many diseases. In contrast with the Darwinian view of nature of selfishness and competence, new holistic approaches are rising. Under these views, the reconstitution of the microbiome comes out as a fundamental therapy for emerging diseases related to biome depletion.

  3. Depleted uranium management alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Hertzler, T.J.; Nishimoto, D.D.

    1994-08-01

    This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process.

  4. Experimental interaction of magma and “dirty” coolants

    Science.gov (United States)

    Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

    2011-03-01

    The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with coolants with >~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate that even modest concentrations of sediment in water will significantly limit heat transfer during non-explosive magma-water interactions. At high concentrations, the dramatic reduction in cooling efficiency

  5. MHD considerations for poloidal-toroidal coolant ducts of self-cooled blankets

    International Nuclear Information System (INIS)

    Hua, T.Q.; Walker, J.S.

    1990-01-01

    Magnetohydrodynamic flows of liquid metals through sharp elbow ducts with rectangular cross sections and with thin conducting walls in the presence of strong uniform magnetic fields are examined. The geometries simulate the poloidaltoroidal coolant channels in fusion tokamak blankets. Analysis for obtaining the three-dimensional numerical solutions are described. Results for pressure drop, velocity profiles and flow distribution are predicted for the upcoming joint ANL/KfK sharp elbow experiment. Results from a parametric study using fusion relevant parameters to investigate the three-dimensional pressure drop are presented for possible applications to blanket designs. 10 refs., 9 refs

  6. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  7. Analyses of Decrease in Reactor Coolant Flow Rate in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

    2011-01-01

    SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

  8. Riddle of depleted uranium

    International Nuclear Information System (INIS)

    Hussein, A.S.

    2005-01-01

    Depleted Uranium (DU) is the waste product of uranium enrichment from the manufacturing of fuel rods for nuclear reactors in nuclear power plants and nuclear power ships. DU may also results from the reprocessing of spent nuclear reactor fuel. Potentially DU has both chemical and radiological toxicity with two important targets organs being the kidney and the lungs. DU is made into a metal and, due to its availability, low price, high specific weight, density and melting point as well as its pyrophoricity; it has a wide range of civilian and military applications. Due to the use of DU over the recent years, there appeared in some press on health hazards that are alleged to be due to DU. In these paper properties, applications, potential environmental and health effects of DU are briefly reviewed

  9. Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Clauzel, M.; Guillodo, M.; Foucault, M. [AREVA NP SAS, Technical Centre, Le Creusot (France); Engler, N.; Chahma, F.; Brun, C. [AREVA NP SAS, Chemistry and Radiochemistry Group, Paris La Defense (France)

    2010-07-01

    The mastering of the reactor coolant system radioactive contamination is a real stake of performance for operating plants and new builds. The reduction of activated corrosion products deposited on RCS surfaces allows minimizing the global dose integrated by workers which supports the ALARA approach. Moreover, the contamination mastering limits the volumic activities in the primary coolant and thus optimizes the reactor shutdown duration and environment releases. The main contamination sources on PWR are due to Co-60 and Co-58 nuclides which come respectively Co-59 and Ni-58, naturally present in alloys used in the RCS. Co is naturally present as an impurity in alloys or as the main component of hardfacing materials (Stellites™). Ni is released mainly by SG tubes which represent the most important surface of the RCS. PWR steam generators (SG), due to the huge wetted surface are the main source of corrosion products release in the primary coolant circuit. As corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup, it is of primary importance to gain a better understanding of phenomenon leading to corrosion product release from SG tubes before setting up mitigation measures. Previous studies have shown that SG tubing made of the same material had different release rates. To find the origin of these discrepancies, investigations have been performed on tubes at the as-received state and after exposure to a nominal primary chemistry in titanium recirculating loop. These investigations highlighted the existence of a correlation between the inner surface metallurgical properties and the release of corrosion products in primary coolant. Oxide films formed in nominal primary chemistry are always protective, their morphology and their composition depending strongly on the geometrical, metallurgical and physico-chemical state of the surface on which they

  10. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

    Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

  11. Estimates of limits to fuel/coolant mixing

    International Nuclear Information System (INIS)

    Park, G.C.; Corradini, M.L.

    1991-06-01

    The vapor explosion process can involve the mixing of fuel with coolant prior to the explosion. A number of analysts have suggested limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a simplified approach is suggested to estimate a limit on the amount of fuel/coolant mixing possible. The approach uses concepts first advanced by Fauske in a different way. The results indicate that fuel temperature, ambient pressure and in particular the mixing length scale D mix are important parameters. For large values of D mix the fuel mass mixed in-vessel is limited to the range of 1-12 metric tons (1-10% of the core mass). (orig.) [de

  12. LOFT advanced densitometer for nuclear loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

    1979-01-01

    A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

  13. A study on the characteristics of alternative coolants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Ji Young; Kim, B. H.; Kim, T. J.; Jeong, K. C.; Choi, Y. D.; Choi, J. H.; Hwang, S. T

    2000-12-01

    The role of the coolant in liquid metal fast breeder reactor is very important for reasons of system safety. Recently, it has revealed that lead and lead-bismuth alloy show good safety characteristics as a fast reactor coolant compared to the sodium, such as low chemical activity, high boiling temperature and more negative void coefficient. So many countries take interest in these metals. The objectives of this project are to study the characteristics of heavy liquid metals(lead, lead-bismuth alloy) and to provide valuble information useful for the estimate the possibilities of its as the alternative coolant materials. An intensive research was performed into the global development status, basic properties, safety assurance methods, and direction of research in the futures and so on.

  14. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  15. The Toxicity of Depleted Uranium

    OpenAIRE

    Briner, Wayne

    2010-01-01

    Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a c...

  16. Integrated Fuel-Coolant Interaction (IFCI 6.0) code

    International Nuclear Information System (INIS)

    Davis, F.J.; Young, M.F.

    1994-04-01

    The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

  17. Estimation of fuel temperature increase during coolant boiloff accident

    International Nuclear Information System (INIS)

    Abe, Kiyoharu

    1981-10-01

    Fuel rod temperature increase during coolant boiloff accident due to unavailable ECCS was analyzed using a simple time dependent model. A standard case was first selected and its results clarified how is the fuel temperature behavior during the boiloff accident. Then the sensitivity studies for various parameters were performed to know what parameters have important roles. As a result of analyses, it was shown that the coolant mixture level in the core has a dominant effect on the core heatup and that fuel rod claddings will probably slump or melt before its full oxidation. (author)

  18. Probabilistic analyses of failure in reactor coolant piping

    International Nuclear Information System (INIS)

    Holman, G.S.

    1984-01-01

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

  19. Sensitivity calculation of the coolant temperature regarding the thermohydraulic parameters

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de; Silva, F.C. da; Thome Filho, Z.D.; Alvim, A.C.M.; Oliveira Barroso, A.C. de.

    1985-01-01

    It's studied the application of the Generalized Perturbation Theory (GPT) in the sensitivity calculation of thermalhydraulic problems, aiming at verifying the viability of the extension of the method. For this, the axial distribution, transient, of the coolant temperature in a PWR channel are considered. Perturbation expressions are developed using the GPT formalism, and a computer code (Tempera) is written, to calculate the channel temperature distribution and the associated importance function, as well as the effect of the thermalhydraulic parameters variations in the coolant temperature (sensitivity calculation). The results are compared with those from the direct calculation. (E.G.) [pt

  20. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  1. Too Depleted to Try? Testing the Process Model of Ego Depletion in the Context of Unhealthy Snack Consumption.

    Science.gov (United States)

    Haynes, Ashleigh; Kemps, Eva; Moffitt, Robyn

    2016-11-01

    The process model proposes that the ego depletion effect is due to (a) an increase in motivation toward indulgence, and (b) a decrease in motivation to control behaviour following an initial act of self-control. In contrast, the reflective-impulsive model predicts that ego depletion results in behaviour that is more consistent with desires, and less consistent with motivations, rather than influencing the strength of desires and motivations. The current study sought to test these alternative accounts of the relationships between ego depletion, motivation, desire, and self-control. One hundred and fifty-six undergraduate women were randomised to complete a depleting e-crossing task or a non-depleting task, followed by a lab-based measure of snack intake, and self-report measures of motivation and desire strength. In partial support of the process model, ego depletion was related to higher intake, but only indirectly via the influence of lowered motivation. Motivation was more strongly predictive of intake for those in the non-depletion condition, providing partial support for the reflective-impulsive model. Ego depletion did not affect desire, nor did depletion moderate the effect of desire on intake, indicating that desire may be an appropriate target for reducing unhealthy behaviour across situations where self-control resources vary. © 2016 The International Association of Applied Psychology.

  2. Method for removing cesium from aqueous liquid, method for purifying the reactor coolant in boiling water and pressurized water reactors and a mixed ion exchanged resin bed, useful in said purification

    International Nuclear Information System (INIS)

    Otte, J.N.A.; Liebmann, D.

    1989-01-01

    The invention relates to a method for removing cesium from an aqueous liquid, and to a resin bed containing a mixture of an anion exchange resin and cation exchange resin useful in said purification. In a preferred embodiment, the present invention is a method for purifying the reactor coolant of a presurized water or boiling water reactor. Said method, which is particularly advantageously employed in purifying the reactor coolant in the primary circuit of a pressurized reactor, comprises contacting at least a portion of the reactor coolant with a strong base anion exchange resin and the strong acid cation exchange resin derived from a highly cross-linked, macroporous copolymer of a monovinylidene aromatic and a cross-linking monomer copolymerizable therewith. Although the reactor coolant can sequentially be contacted with one resin type and thereafter with the second resin type, the contact is preferably conducted using a resin bed comprising a mixture of the cation and anion exchange resins. 1 fig., refs

  3. Directly connected heat exchanger tube section and coolant-cooled structure

    Science.gov (United States)

    Chainer, Timothy J; Coico, Patrick A; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Schmidt, Roger R; Steinke, Mark E

    2014-04-01

    A cooling apparatus for an electronics rack is provided which includes an air-to-liquid heat exchanger, one or more coolant-cooled structures and a tube. The heat exchanger, which is associated with the electronics rack and disposed to cool air passing through the rack, includes a plurality of distinct, coolant-carrying tube sections, each tube section having a coolant inlet and a coolant outlet, one of which is coupled in fluid communication with a coolant loop to facilitate flow of coolant through the tube section. The coolant-cooled structure(s) is in thermal contact with an electronic component(s) of the rack, and facilitates transfer of heat from the component(s) to the coolant. The tube connects in fluid communication one coolant-cooled structure and the other of the coolant inlet or outlet of the one tube section, and facilitates flow of coolant directly between that coolant-carrying tube section of the heat exchanger and the coolant-cooled structure.

  4. Depletable resources and the economy

    NARCIS (Netherlands)

    Heijman, W.J.M.

    1991-01-01

    The subject of this thesis is the depletion of scarce resources. The main question to be answered is how to avoid future resource crises. After dealing with the complex relation between nature and economics, three important concepts in relation with resource depletion are discussed: steady state,

  5. The Toxicity of Depleted Uranium

    Directory of Open Access Journals (Sweden)

    Wayne Briner

    2010-01-01

    Full Text Available Depleted uranium (DU is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a clear and defined set of symptoms. Chronic low-dose, or subacute, exposure to depleted uranium alters the appearance of milestones in developing organisms. Adult animals that were exposed to depleted uranium during development display persistent alterations in behavior, even after cessation of depleted uranium exposure. Adult animals exposed to depleted uranium demonstrate altered behaviors and a variety of alterations to brain chemistry. Despite its reduced level of radioactivity evidence continues to accumulate that depleted uranium, if ingested, may pose a radiologic hazard. The current state of knowledge concerning DU is discussed.

  6. Numerical experimentation on convective coolant flow in Ghana ...

    African Journals Online (AJOL)

    Numerical experiments on one dimensional convective coolant flow during steady state operation of the Ghana Research Reactor-1 (GHARR-I) were performed to determine the thermal hydraulic parameters of temperature, density and flow rate. The computational domain was the reactor vessel, including the reactor core.

  7. Reactor coolant and associated systems in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

  8. Stress analysis of a LWR main coolant pump

    International Nuclear Information System (INIS)

    Novobilsky, R.

    1985-01-01

    Because of its importance for the safe operation of a nuclear power plant the main coolant pump of a LWR has to be designed according to the same rules as the pressure vessel or the main coolant piping itself. As prescribed in the design specification, the stress and fatigue analysis of the main coolant pump has to follow the rules of the ASME Code as well as the German KTA-Regel. The main coolant pump is designed as an axial-flow pump. The housing and the housing cover are made of forged steel. At the inner surface, the housing has a 6 mm thick austenitic cladding. The pump operation details are given. The finite element model of the pump was constructed using the computer code PDA/PATRAN-G. Because of the symmetry, only one half of the housing has to be modelled. The stress analysis of the pump housing was performed using MSC/NASTRAN. Details are given. The results are discussed. (U.K.)

  9. En-mass coolant channel replacement in Indian PHWR

    International Nuclear Information System (INIS)

    Varghese, Joy P.; Verma, V.S.; Rajput, C.D.; Ramamurthy, K.

    2010-01-01

    This paper describes in brief the need for coolant channel replacement, challenges faced, methodology adopted to successfully overcome the shortcomings, improvements made in various areas that set a new standard and lifted the Indian image in the global nuclear industry. These changes have streamlined the EMCCR programme, which was followed on during the subsequent campaigns

  10. Observed and simulated depletion layers with southward IMF

    Directory of Open Access Journals (Sweden)

    N. C. Maynard

    2004-06-01

    Full Text Available We present observations from the Polar satellite that confirm the existence of two types of depletion layers predicted under southward interplanetary magnetic field (IMF conditions in magnetohydrodynamic simulations. The first depletion type occurs along the stagnation line when IMF BX and/or dipole tilt are/is present. Magnetic merging occurred away from the equator (Maynard et al., 2003 and flux pile-ups developed while the field lines drape to the high-latitude merging sites. This high-shear type of depletion is consistent with the depletion layer model suggested by Zwan and Wolf (1976 for low-shear northward IMF conditions. Expected sites for depletion layers are associated with places where IMF tubes of force first impinge upon the magnetopause. The second depletion type develops poleward of the cusp. Under strongly driven conditions, magnetic fields from Region 1 current closure over the lobes (Siscoe et al., 2002c cause the high-latitude magnetopause to bulge outward, creating a shoulder above the cusp. These shoulders present the initial obstacle with which the IMF interacts. Flow is impeded, causing local flux pile-ups and low-shear depletion layers to form poleward of the cusps. Merging at the high-shear dayside magnetopause is consequently delayed. In both low- and high-shear cases, we show that the depletion layer structure is part of a slow mode wave standing in front of the magnetopause. As suggested by Southwood and Kivelson (1995, the depletions are rarefactions on the magnetopause side of slow-mode density compressions. While highly sheared magnetic fields are often used as proxies for ongoing local magnetic merging, depletion layers are prohibited at merging locations. Therefore, the existence of a depletion layer is evidence that the location of merging must be remote relative to the observation.

  11. Observed and simulated depletion layers with southward IMF

    Directory of Open Access Journals (Sweden)

    N. C. Maynard

    2004-06-01

    Full Text Available We present observations from the Polar satellite that confirm the existence of two types of depletion layers predicted under southward interplanetary magnetic field (IMF conditions in magnetohydrodynamic simulations. The first depletion type occurs along the stagnation line when IMF BX and/or dipole tilt are/is present. Magnetic merging occurred away from the equator (Maynard et al., 2003 and flux pile-ups developed while the field lines drape to the high-latitude merging sites. This high-shear type of depletion is consistent with the depletion layer model suggested by Zwan and Wolf (1976 for low-shear northward IMF conditions. Expected sites for depletion layers are associated with places where IMF tubes of force first impinge upon the magnetopause. The second depletion type develops poleward of the cusp. Under strongly driven conditions, magnetic fields from Region 1 current closure over the lobes (Siscoe et al., 2002c cause the high-latitude magnetopause to bulge outward, creating a shoulder above the cusp. These shoulders present the initial obstacle with which the IMF interacts. Flow is impeded, causing local flux pile-ups and low-shear depletion layers to form poleward of the cusps. Merging at the high-shear dayside magnetopause is consequently delayed. In both low- and high-shear cases, we show that the depletion layer structure is part of a slow mode wave standing in front of the magnetopause. As suggested by Southwood and Kivelson (1995, the depletions are rarefactions on the magnetopause side of slow-mode density compressions. While highly sheared magnetic fields are often used as proxies for ongoing local magnetic merging, depletion layers are prohibited at merging locations. Therefore, the existence of a depletion layer is evidence that the location of merging must be remote relative to the observation.

  12. Problems of heavy coolant technology for accelerator-driven systems

    International Nuclear Information System (INIS)

    Gromov, B.F.; Orlov, Yu.I.; Gulevskij, V.A.; Martinov, P.N.; Efimov, E.I.

    1999-01-01

    The title topic is discussed, and the activities of nuclides in the target gas system by 1 MW power after 6 months of work are tabulated. The following facts are stressed: Realization of traditional technology based mainly on cover gas replacement, release and feeding of gas mixtures into the loop is complicated and needs developing and applying devices for purification of the gases removed and radiation shielding of these devices at least. Such developments are especially urgent for ADS projects where coolant circulation is intensified by gas lift. Traditional technology is unable to solve a number of problems associated with the elimination of electropositive impurities (Au, Ag, Hg and others) that make a significant contribution to change compound and radio activity of ADS coolants. The task to remove such impurities was not set for traditional nuclear power plants, so that appropriate developments began only recently. New processes that could replace the developed processes have not been revealed in the course of investigations, however there is an opportunity of using new methods of their realization. For example, the processes of purification and regulation of coolant quality using chemically active gases can be implemented without using external gas systems. The solid hydride, metallooxide and other chemical compounds located in the gas volume or in the coolant loops of ADS can be the sources and absorbents of chemically active gas components. Applying to ADS conditions the emphasis of some processes goals can be slightly changed. For example: regulation of dissolved oxygen activity can be performed as well (probably, mainly) for oxygen refining of coolant by means of electronegative impurities transition into solid-phase oxide compounds; gas and coolant filtering can also be performed to localize radioactive impurities in the ADS circuit. Probably, the coolant technology can be employed to solve a number of tasks: to localize radioactive impurities in the

  13. It is chloride depletion alkalosis, not contraction alkalosis.

    Science.gov (United States)

    Luke, Robert G; Galla, John H

    2012-02-01

    Maintenance of metabolic alkalosis generated by chloride depletion is often attributed to volume contraction. In balance and clearance studies in rats and humans, we showed that chloride repletion in the face of persisting alkali loading, volume contraction, and potassium and sodium depletion completely corrects alkalosis by a renal mechanism. Nephron segment studies strongly suggest the corrective response is orchestrated in the collecting duct, which has several transporters integral to acid-base regulation, the most important of which is pendrin, a luminal Cl/HCO(3)(-) exchanger. Chloride depletion alkalosis should replace the notion of contraction alkalosis.

  14. Coolant voiding analysis following SGTR for an HLMC reactor

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

    2000-01-01

    Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

  15. Examination of disks from the IPNS depleted uranium target

    International Nuclear Information System (INIS)

    Strain, R.V.; Carpenter, J.M.

    1995-10-01

    This report describes the results of examining the Zircaloy-2 clad depleted uranium disks from the Intense Pulse Neutron Source (IPNS) Target. That target operated from August, 1981 to June, 1988 and from September, 1991 to September, 1992 at 450 MeV, pulsing at 30 Hz with a time average proton current of about 15 microA. The target was removed from service when the presence of fission products ( 135 Xe) in the coolant cover gas indicated a failure in the Zircaloy-2 cladding. Altogether, the target had absorbed about 240 mA hours of proton current, and endured between 50,000 and 100,000 thermal cycles. The purpose of the examination was to assess the condition of the disks and determine the cause of the cladding failure. The results of visual, gamma ray scanning, and destructive metallurgical examination of two disks are described

  16. Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

    International Nuclear Information System (INIS)

    Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

    2010-01-01

    Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

  17. Coolant mixing in the LMFBR outlet plenum

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-06-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds Number (Re) values of 33000 and 70000 in a 1/15-scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet velocity field, upon the degree of inlet turbulence, and upon the turbulence momentum exchange model used in the calculations. It is found in the FFTF geometry that the TEACH-T predictions are better than that of VARR-II, and in the CRBR geometry neither code provides a good prediction of the observed behavior. From the sensitivity analysis, it is found that the production and dissipation of turbulence are the dominant terms in the transport equations for turbulent kinetic energy and turbulent energy dissipation rate, and the diffusion terms are relatively small. From the same study a new set of empirical constants for the turbulence model is evolved for the prediction of plenum flows

  18. Cross-comparison of fast reactor concepts with various coolants

    International Nuclear Information System (INIS)

    Hejzlar, Pavel; Todreas, Neil E.; Shwageraus, Eugene; Nikiforova, Anna; Petroski, Robert; Driscoll, Michael J.

    2009-01-01

    Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl 2 (LSFR), sodium (SFR), and supercritical CO 2 (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO 2 power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO 2 reflector to achieve negative coolant void worth-one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.

  19. Laser-based sensor for a coolant leak detection in a nuclear reactor

    Science.gov (United States)

    Kim, T.-S.; Park, H.; Ko, K.; Lim, G.; Cha, Y.-H.; Han, J.; Jeong, D.-Y.

    2010-08-01

    Currently, the nuclear industry needs strongly a reliable detection system to continuously monitor a coolant leak during a normal operation of reactors for the ensurance of nuclear safety. In this work, we propose a new device for the coolant leak detection based on tunable diode laser spectroscopy (TDLS) by using a compact diode laser. For the feasibility experiment, we established an experimental setup consisted of a near-IR diode laser with a wavelength of about 1392 nm, a home-made multi-pass cell and a sample injection system. The feasibility test was performed for the detection of the heavy water (D2O) leaks which can happen in a pressurized heavy water reactor (PWHR). As a result, the device based on the TDLS is shown to be operated successfully in detecting a HDO molecule, which is generated from the leaked heavy water by an isotope exchange reaction between D2O and H2O. Additionally, it is suggested that the performance of the new device, such as sensitivity and stability, can be improved by adapting a cavity enhanced absorption spectroscopy and a compact DFB diode laser. We presume that this laser-based leak detector has several advantages over the conventional techniques currently employed in the nuclear power plant, such as radiation monitoring, humidity monitoring and FT-IR spectroscopy.

  20. Hsp90 depletion goes wild

    OpenAIRE

    Siegal, Mark L; Masel, Joanna

    2012-01-01

    Abstract Hsp90 reveals phenotypic variation in the laboratory, but is Hsp90 depletion important in the wild? Recent work from Chen and Wagner in BMC Evolutionary Biology has discovered a naturally occurring Drosophila allele that downregulates Hsp90, creating sensitivity to cryptic genetic variation. Laboratory studies suggest that the exact magnitude of Hsp90 downregulation is important. Extreme Hsp90 depletion might reactivate transposable elements and/or induce aneuploidy, in addition to r...

  1. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  2. Nonlinear dynamic analysis of nuclear reactor primary coolant systems

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

    1979-01-01

    The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

  3. Nuclear reactor coolant pump impeller/shaft assembly

    International Nuclear Information System (INIS)

    Jenkins, L.S.

    1987-01-01

    A pump is described comprising: (a) a casing having an inlet and an outlet in fluid communication for circulating fluid coolant through the pump; (b) a shaft positioned in the casing; (c) an impeller nut connected to the shaft; (d) a lockbolt fixedly connecting the impeller nut relative to the shaft; (e) passageway means with first and second ends for directing fluid from the lockbolt to the shaft; (f) first conduit means formed in the impeller nut in fluid communication with the first end of the passageway means and the inlet of the casing; and (g) second conduit means formed in the impeller nut in fluid communication with the second end of the passageway means and the outlet of the casing. A portion of the fluid coolant circulating through the inlet of the casing is pumped through the first conduit means, through the passageway means, out the second conduit means and into the outlet of the casing

  4. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  5. Peculiarities of evolution of shock waves generated by boiling coolant

    Science.gov (United States)

    Alekseev, M. V.; Vozhakov, I. S.; Lezhnin, S. I.; Pribaturin, N. A.

    2016-11-01

    Simulation of compression wave generation and evolution at the disk target was performed for the case of explosive-type boiling of coolant; the boiling is initiated by endwall rupture of a high-pressure pipeline. The calculations were performed for shock wave amplitude at different times and modes of pipe rupture. The simulated pressure of a target-reflected shock wave is different from the theoretical value for ideal gas; this discrepancy between simulation and theory becomes lower at higher distances of flow from the nozzle exit. Comparative simulation study was performed for flow of two-phase coolant with account for slip flow effect and for different sizes of droplets. Simulation gave the limiting droplet size when the single-velocity homogeneous flow model is valid, i.e., the slip flow effect is insignificant.

  6. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

  7. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  8. Method of injecting iron ion into reactor coolant

    International Nuclear Information System (INIS)

    Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

    1988-01-01

    Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

  9. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  10. Optimum parameters for treating coolant wastewater using PVDF-membrane

    Directory of Open Access Journals (Sweden)

    Yuliwati Erna

    2018-01-01

    Full Text Available Chemical oxygen demand (COD removal on ultrafiltration membrane were studied from treating of coolant wastewater. The membranes were fabricated via the dry-jet wet spinning method. The experiment was conducted using synthetic coolant wastewater as effluent, and an experimental set-up consist of membrane reservoir was used throughout the investigation. Deposition and accumulation of suspended solids on the membrane surface during filtration were prohibited with continuous aeration. Membrane performance was measured as well as flux and COD removal efficiency. The RSM was performed to investigate the optimised process conditions for treating coolant wastewater. Results using RSM-approximation have demonstrated the improvement in COD removal of 99.63%. This improvement is achieved by optimised process conditions of mixed liquor suspended solids (MLSS concentration, air bubble flow rate (ABFR, hydraulic retention time (HRT and pH at 6.00 g/L, 3.05 ml/min, 280.50 min, and 6.50 respectively.

  11. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  12. On-line real time gamma analysis of primary coolant

    International Nuclear Information System (INIS)

    Kalechstein, W.; Kupca, S.; Lipsett, J.J.

    1985-10-01

    The evolution of failed fuel monitoring at CANDU power stations is briefly summarized and the design of the latest system for failed fuel detection at a multi-unit power station is described. At each reactor, the system employs a germanium spectrometer combined with a novel spectrum analyzer that simultaneously accumulates the gamma-ray spectrum of the coolant and provides the control room with the concentration of radioisotope activity in the coolant for the gaseous fission products Xe-133, Xe-135, Kr-88 and I-131 in real time and with statistical precision independent of count rate. A gross gamma monitor is included to provide independent information on the level of radioactivity in the coolant and extend the measurement range at very high count rates. A central computer system archives spectra received from all four spectrum analyzers and provides both the activity concentrations and the release rates of specified isotopes. Compared with previous systems the current design offers improvements in that the activity concentrations are updated much more frequently, improved tools are provided for long term surveillance of the heat transport system and the monitor is more reliable and less costly

  13. Coolant Design System for Liquid Propellant Aerospike Engines

    Science.gov (United States)

    McConnell, Miranda; Branam, Richard

    2015-11-01

    Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

  14. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  15. Summer time Fe depletion in the Antarctic mesopause region

    Science.gov (United States)

    Viehl, T. P.; Höffner, J.; Lübken, F.-J.; Plane, J. M. C.; Kaifler, B.; Morris, R. J.

    2015-05-01

    We report common volume measurements of Fe densities, temperatures and ice particle occurrence in the mesopause region at Davis Station, Antarctica (69°S) in the years 2011-2012. Our observations show a strong correlation of the Fe-layer summer time depletion with temperature, but no clear causal relation with the onset or occurrence of ice particles measured as noctilucent clouds (NLC) or polar mesosphere summer echoes (PMSE). The combination of these measurements indicates that the strong summer depletion can be explained by gas-phase chemistry alone and does not require heterogeneous removal of Fe and its compounds on ice particles.

  16. Analysis of actual status of works on technology of heavy liquid metal coolants

    International Nuclear Information System (INIS)

    Martynov, P.N.; Askhadullin, R.Sh.; Orlov, Yu.I.; Storozhenko, A.N.

    2014-01-01

    Principle duties in heavy liquid metal coolant technology (HLMC) are provision of the purity of coolant and surfaces of circulation loop for maintenance of design thermohydraulic characteristics, prevention of structural materials corrosion and erosion during long service life and present-day safety precautions on different stages of reactor facility operation. For this reason, current HLMC (Pb-Bi, Pb) technology must include coolant pre-operation and charging; monitoring and regulating of coolant oxygen potential; hydrogen purification of coolant and surfaces of circulation loop from lead oxides-based slags; coolant filtration; reactor cover gas purification from coolant aerosols. The current topical problem is personnel training on the questions of HLMC technology [ru

  17. A comparison of core perturbation by coolant loss between sodium and lead-bismuth cooled reactor

    International Nuclear Information System (INIS)

    Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

    2003-01-01

    This study performs a comparative analysis of the core perturbation caused by coolant loss between sodium and lead-bismuth eutectic. Considering the Zr-based and the U-based fuel in a 1,000MWth class reactor for TRU incineration, we investigate which coolant shows better performance for negative coolant loss reactivity in each case of fuel type. The calculation results show that in the case of U-based fuel, sodium gives rise to more positive coolant loss reactivity than lead-bismuth. However, when the Zr-based (U-free) fuel is considered, sodium offers negative coolant loss reactivity, whereas lead-bismuth makes the coolant loss reactivity positive. It is recommended to employ sodium coolant for the fertile-free fueled core and lead-bismuth for the core with fertile nuclides

  18. BWR loss-of-coolant accident tests at ROSA-III with high temperature emergency core coolant injection

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Kukita, Yutaka; Tasaka, Kanji

    1988-01-01

    The effects of emergency core coolant (ECC) temperature on the performance of the emergency core cooling system (ECCS) during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) were investigated experimentally using the Rig of Safety Assessment (ROSA)-3 integral test facility. The ECC temperature had no direct influence on the ECCS core cooling performance, since ECC became nearly saturated before reaching the core irrespective of the initial temperature, however, had indirect effects by changing the vessel pressure response. The ECCS injection timings and flow rates, and the core inlet flooding behavior were affected. The measured peak cladding temperature (PCT) was not affected by the ECC temperature for both large (200%) and small (5%) break tests. (author)

  19. Feeding and purge systems of coolant primary circuit and coolant secondary circuit control of the I sup(123) target

    International Nuclear Information System (INIS)

    Almeida, G.L. de.

    1986-01-01

    The Radiation Protection Service of IEN (Brazilian-CNEN) detected three faults in sup(123)I target cooling system during operation process for producing sup(123)I: a) non hermetic vessel containing contaminated water from primary coolant circuit; possibility of increasing radioactivity in the vessel due to accumulation of contaminators in cooling water and; situation in region used for personnels to arrange and adjust equipments in nuclear physics area, to carried out maintenance of cyclotron and target coupling in irradiation room. The primary circuit was changed by secondary circuit for target coolant circulating through coil of tank, which receive weater from secondary circuit. This solution solved the three problems simultaneously. (M.C.K.)

  20. Membrane technology for treating of waste nanofluids coolant: A review

    Science.gov (United States)

    Mohruni, Amrifan Saladin; Yuliwati, Erna; Sharif, Safian; Ismail, Ahmad Fauzi

    2017-09-01

    The treatment of cutting fluids wastes concerns a big number of industries, especially from the machining operations to foster environmental sustainability. Discharging cutting fluids, waste through separation technique could protect the environment and also human health in general. Several methods for the separation emulsified oils or oily wastewater have been proposed as three common methods, namely chemical, physicochemical and mechanical and membrane technology application. Membranes are used into separate and concentrate the pollutants in oily wastewater through its perm-selectivity. Meanwhile, the desire to compensate for the shortcomings of the cutting fluid media in a metal cutting operation led to introduce the using of nanofluids (NFs) in the minimum quantity lubricant (MQL) technique. NFs are prepared based on nanofluids technology by dispersing nanoparticles (NPs) in liquids. These fluids have potentially played to enhance the performance of traditional heat transfer fluids. Few researchers have studied investigation of the physical-chemical, thermo-physical and heat transfer characteristics of NFs for heat transfer applications. The use of minimum quantity lubrication (MQL) technique by NFs application is developed in many metal cutting operations. MQL did not only serve as a better alternative to flood cooling during machining operation and also increases better-finished surface, reduces impact loads on the environment and fosters environmental sustainability. Waste coolant filtration from cutting tools using membrane was treated by the pretreated process, coagulation technique and membrane filtration. Nanomaterials are also applied to modify the membrane structure and morphology. Polyvinylidene fluoride (PVDF) is the better choice in coolant wastewater treatment due to its hydrophobicity. Using of polyamide nanofiltration membranes BM-20D and UF-PS-100-100, 000, it resulted in the increase of permeability of waste coolant filtration. Titanium dioxide

  1. Hsp90 depletion goes wild

    Directory of Open Access Journals (Sweden)

    Siegal Mark L

    2012-02-01

    Full Text Available Abstract Hsp90 reveals phenotypic variation in the laboratory, but is Hsp90 depletion important in the wild? Recent work from Chen and Wagner in BMC Evolutionary Biology has discovered a naturally occurring Drosophila allele that downregulates Hsp90, creating sensitivity to cryptic genetic variation. Laboratory studies suggest that the exact magnitude of Hsp90 downregulation is important. Extreme Hsp90 depletion might reactivate transposable elements and/or induce aneuploidy, in addition to revealing cryptic genetic variation. See research article http://wwww.biomedcentral.com/1471-2148/12/25

  2. Analysis of the VVER-1000 coolant transient benchmark phase 1 with the code system RELAP5/PARCS

    International Nuclear Information System (INIS)

    Victor Hugo Sanchez Espinoza

    2005-01-01

    the test and its effects on the local power distribution within the core. The code's predictions are strongly influenced by the way how coolant mixing is modeled by the analyst by means of 1D-thermal hydraulic codes. Sensitivity evaluations are therefore necessary to identify the most important phenomena and assumptions affecting the numerical predictions of coupled codes. Selected results of these investigations will be presented and discussed. A comparison of the thermal hydraulic data obtained during the tests with the code's predictions will be also given. It can be stated that even though the overall trends of most plant parameters are in a reasonable agreement with the experimental data. These investigations show also that multidimensional thermal hydraulic models are needed for a more realistic description of the coolant mixing phenomena within the reactor pressure vessel. Hence the subsequent Phase 2 of this V1000- CT benchmark is focused on CFD-based simulations for transient conditions typical of a main steam line break transients of VVER-1000 reactors. (author)

  3. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    International Nuclear Information System (INIS)

    Sanchez-Espinoza, Victor Hugo

    2008-07-01

    local power distribution within the core. The code's predictions are strongly influenced by the way how coolant mixing is modeled by the analyst by means of 1D-thermal hydraulic codes. Sensitivity evaluations are therefore necessary to identify the most important phenomena and assumptions affecting the numerical predictions of coupled codes. Selected results of these investigations will be presented and discussed. A comparison of the thermal hydraulic data obtained during the tests with the code's predictions will be also given. It can be stated that even though the overall trends of most plant parameters are in a reasonable agreement with the experimental data, these investigations show that multidimensional thermal hydraulic models are needed for a more realistic description of the coolant mixing phenomena within the reactor pressure vessel. Hence the subsequent Phase 2 of this V1000-CT benchmark is focused on CFD-based simulations for transient conditions typical of a main steam line break transients of VVER-1000 reactors. (orig.)

  4. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    local power distribution within the core. The code's predictions are strongly influenced by the way how coolant mixing is modeled by the analyst by means of 1D-thermal hydraulic codes. Sensitivity evaluations are therefore necessary to identify the most important phenomena and assumptions affecting the numerical predictions of coupled codes. Selected results of these investigations will be presented and discussed. A comparison of the thermal hydraulic data obtained during the tests with the code's predictions will be also given. It can be stated that even though the overall trends of most plant parameters are in a reasonable agreement with the experimental data, these investigations show that multidimensional thermal hydraulic models are needed for a more realistic description of the coolant mixing phenomena within the reactor pressure vessel. Hence the subsequent Phase 2 of this V1000-CT benchmark is focused on CFD-based simulations for transient conditions typical of a main steam line break transients of VVER-1000 reactors. (orig.)

  5. PUMP: analog-hybrid reactor coolant hydraulic transient model

    International Nuclear Information System (INIS)

    Grandia, M.R.

    1976-03-01

    The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

  6. Vegetable oils: liquid coolants for solar heating and cooling applications

    Energy Technology Data Exchange (ETDEWEB)

    Ingley, H A

    1980-02-01

    It has been proposed that vegetable oils, renewable byproducts of agriculture processes, be investigated for possible use as liquid coolants. The major thrust of the project was to investigate several thermophysical properties of the four vegetable oils selected. Vapor pressures, specific heat, viscosity, density, and thermal conductivity were determined over a range of temperatures for corn, soybean, peanut, and cottonseed oil. ASTM standard methods were used for these determinations. In addition, chemical analyses were performed on samples of each oil. The samples were collected before and after each experiment so that any changes in composition could be noted. The tests included iodine number, fatty acid, and moisture content determination. (MHR)

  7. Use of flow models to analyse loss of coolant accidents

    International Nuclear Information System (INIS)

    Pinet, Bernard

    1978-01-01

    This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

  8. COPDIRC - calculation of particle deposition in reactor coolants

    International Nuclear Information System (INIS)

    Reeks, M.W.

    1982-06-01

    A description is given of a computer code COPDIRC intended for the calculation of the deposition of particulate onto smooth perfectly sticky surfaces in a gas cooled reactor coolant. The deposition is assumed to be limited by transport in the boundary layer adjacent to the depositing surface. This implies that the deposition velocity normalised with respect to the local friction velocity, is an almost universal function of the normalised particle relaxation time. Deposition is assumed similar to deposition in an equivalent smooth perfectly absorbing pipe. The deposition is calculated using 2 models. (author)

  9. Calculational advance in the modeling of fuel-coolant interactions

    International Nuclear Information System (INIS)

    Bohl, W.R.

    1982-01-01

    A new technique is applied to numerically simulate a fuel-coolant interaction. The technique is based on the ability to calculate separate space- and time-dependent velocities for each of the participating components. In the limiting case of a vapor explosion, this framework allows calculation of the pre-mixing phase of film boiling and interpenetration of the working fluid by hot liquid, which is required for extrapolating from experiments to a reactor hypothetical accident. Qualitative results are compared favorably to published experimental data where an iron-alumina mixture was poured into water. Differing results are predicted with LMFBR materials

  10. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  11. DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

    Directory of Open Access Journals (Sweden)

    KE CHON CHOI

    2013-02-01

    In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129I was examined, as was the effect of 3H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3H presence was found with activity concentrations of 3H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

  12. Definition of loss-of-coolant accident radiation source

    International Nuclear Information System (INIS)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist

  13. Impact of mineral resource depletion

    CSIR Research Space (South Africa)

    Brent, AC

    2006-09-01

    Full Text Available In a letter to the editor, the authors comment on BA Steen's article on "Abiotic Resource Depletion: different perceptions of the problem with mineral deposits" published in the special issue of the International Journal of Life Cycle Assessment...

  14. Design technology development of the main coolant pump for an integral reactor

    International Nuclear Information System (INIS)

    Park, J. S.; Lee, J. S.; Kim, M. H.; Kim, D. W.; Kim, J. I.

    2004-01-01

    All of the reactor coolant pump currently used in commercial nuclear power plant were imported from foreign country. Now, the developing program of design technology for the reactor coolant pump will be started in a few future by domestic researchers. At this stage, the design technology of the main coolant pump for an integral reactor is developed based on the regulation of domestic nuclear power plant facilities. The main coolant pump is a canned motor axial pump, which accommodates all constraints required from the integral reactor system. The main coolant pump does not have mechanical seal device because the rotor of motor and the shaft of impeller are the same one. There is no flywheel on the rotating shaft of main coolant pump so that the coastdown duration time is short when the electricity supply is cut off

  15. Optimization of the water chemistry of the primary coolant at nuclear power plants with VVER

    International Nuclear Information System (INIS)

    Barmin, L. F.; Kruglova, T. K.; Sinitsyn, V. P.

    2005-01-01

    Results of the use of automatic hydrogen-content meter for controlling the parameter of 'hydrogen' in the primary coolant circuit of the Kola nuclear power plant are presented. It is shown that the correlation between the 'hydrogen' parameter in the coolant and the 'hydrazine' parameter in the makeup water can be used for controlling the water chemistry of the primary coolant system, which should make it possible to optimize the water chemistry at different power levels

  16. Experimental Investigation of Coolant Boiling in a Half-Heated Circular Tube - Final CRADA Report

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Wenhua [Argonne National Lab. (ANL), Argonne, IL (United States); Singh, Dileep [Argonne National Lab. (ANL), Argonne, IL (United States); France, David M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-11-01

    Coolant subcooled boiling in the cylinder head regions of heavy-duty vehicle engines is unavoidable at high thermal loads due to high metal temperatures. However, theoretical, numerical, and experimental studies of coolant subcooled flow boiling under these specific application conditions are generally lacking in the engineering literature. The objective of this project was to provide such much-needed information, including the coolant subcooled flow boiling characteristics and the corresponding heat transfer coefficients, through experimental investigations.

  17. Preliminary design of reactor coolant pump canned motor for AC600

    International Nuclear Information System (INIS)

    Deng Shaowen

    1998-01-01

    The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

  18. Pengaruh Penambahan Zat Aditif (Calcium Hypochlorite) terhadap Coolant Properties (Viscosity) Minyak Sawit

    OpenAIRE

    Skp, Buma Gempa; Masnur, Dedy; HS, Irdoni

    2014-01-01

    Cutting fluid (coolant) are required in the machining processes. Most of cutting fluid are from mineral oil, and they are as unrenewable resource, therefore look for alternative resources ofmineral oil, which derived from vegetables oil. Palm oil have potency based of cutting fluid, however of coolant properties increased additive. This research want investigate effect ofadditive on phiysical properties of coolant. The method on mixture combining palm oil and additive (calcium hypochlorite) a...

  19. Fitness for service assessment of coolant channels of Indian PHWRs

    Science.gov (United States)

    Sinha, R. K.; Sinha, S. K.; Madhusoodanan, K.

    2008-12-01

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 × 10 17 n/m 2 s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

  20. Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

    We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

  1. Station blackout with reactor coolant pump seal leakage

    International Nuclear Information System (INIS)

    Evinay, A.

    1993-01-01

    The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, open-quotes Loss of All Alternating Current Power.close quotes The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, open-quotes Station Blackout,close quotes to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate

  2. Coolant clean-up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

    1987-01-01

    Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

  3. Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

    International Nuclear Information System (INIS)

    Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

    2014-01-01

    The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

  4. The effect of coolant boiling on the molten metal pool heat transfer with local solidification

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Y.; Chung, Chang Hyun; Park, Rae joon; Kim, Sang Baik

    2000-01-01

    This study is concerned with the experimental test and numerical analysis of the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. In the test, the metal pool is heated from the bottom surface and coolant is injected into the molten metal pool. Experiments were performed by changing the test section bottom surface temperature of the metal layer and the coolant injection rate. The two-phase boiling coolant experimental results are compared against the dry test without coolant or solidification of the molten metal pool, and against the crust formation experiment with subcooled coolant. Also, a numerical analysis is performed to check on the measured data. The numerical program is developed using the enthalpy method, the finite volume method and the SIMPLER algorithm. The experimental results of the heat transfer show general agreement with the calculated values. The present empirical test and numerical results of the heat transfer on the molten metal pool are apparently higher than those without coolant boiling. This is probably because this experiment was performed in concurrence of solidification in the molten metal pool and the rapid boiling of the coolant. The other experiments were performed without coolant boiling and the correlation was developed for the pure molten metal without phase change. (author)

  5. Test results of the reactor inlet coolant temperature control system of HTTR

    International Nuclear Information System (INIS)

    Saito, Kenji; Nakagawa, Shigeaki; Hirato, Yoji

    2004-04-01

    The reactor control system of HTTR is composed of the reactor power control system, the reactor inlet coolant temperature control system, the primary coolant flow rate control system and so on. The reactor control system of HTTR achieves reactor power 30 MW, reactor outlet coolant temperature 850degC, reactor inlet coolant temperature 395degC under the condition that primary coolant flow rate is fixed. In the Rise-to-Power Test, the performance test of the reactor inlet coolant temperature control system was carried out in order to confirm the control capability of this control system. This report shows the test results of performance test. As a result, the control parameters, which can control the reactor inlet coolant temperature stably during the reactor operation, were successfully selected. And it was confirmed that the reactor inlet coolant temperature control system has the capability of controlling the reactor inlet coolant temperature stably against any disturbances on the basis of operational condition of HTTR. (author)

  6. Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

    International Nuclear Information System (INIS)

    Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

    1976-01-01

    In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

  7. Method for controlling a coolant liquid surface of cooling system instruments in an atomic power plant

    International Nuclear Information System (INIS)

    Monta, Kazuo.

    1974-01-01

    Object: To prevent coolant inventory within a cooling system loop in an atomic power plant from being varied depending on loads thereby relieving restriction of varied speed of coolant flow rate to lowering of a liquid surface due to short in coolant. Structure: Instruments such as a superheater, an evaporator, and the like, which constitute a cooling system loop in an atomic power plant, have a plurality of free liquid surface of coolant. Portions whose liquid surface is controlled and portions whose liquid surface is varied are adjusted in cross-sectional area so that the sum total of variation in coolant inventory in an instrument such as a superheater provided with an annulus portion in the center thereof and an inner cylindrical portion and a down-comer in the side thereof comes equal to that of variation in coolant inventory in an instrument such as an evaporator similar to the superheater. which is provided with an overflow pipe in its inner cylindrical portion or down-comer, thereby minimizing variation in coolant inventory of the entire coolant due to loads thus minimizing variation in varied speed of the coolant. (Kamimura, M.)

  8. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  9. Manometric core liquid level depression during a small-break loss-of-coolant accident in a Westinghouse-type pressurized water reactor

    International Nuclear Information System (INIS)

    Kukita, Y.; Yonomoto, T.; Nakamura, H.; Tasaka, K.

    1988-01-01

    During a certain class of cold leg small-break loss-of-coolant accident (SBLOCA) in a pressurized water reactor (PWR), the primary coolant inventory loss will continue until the break flow rate decreases sufficiently, due to break uncovery or primary depressurization, to allow the emergency core cooling system (ECCS) to make up the coolant loss. Such coolant inventory depletion will involve concurrent liquid level depressions in the cross-over leg downflow side, for each loop, and in the vessel riser section (upper plenum and core). Meanwhile, the cross-over legs upflow side, cold legs and downcomer will remain liquid filled up to the break elevation. The core level depressions will continue until the cross-over leg downflow-side level reaches to the bottom of the legs and thus allows the vapor to clear the liquid seal in the cross-over legs toward the break. When this liquid clearing (loop seal clearing) initiates, the core liquid level starts to recover. This occurs because the vapor can now reach to the downcomer and allows a manometric flow of the downcomer liquid inventory into the core region. This paper presents experimental results on the manometric core level depression phenomenon obtained from a LSTF 5% cold leg break test, Run SB-CL-08. The SG primary side behavior in this test, including non-uniform parallel channel behavior of U-tubes, is described in Ref. 7. The test results are described in comparison with post-test analysis results obtained with the RELAP5/MOD2 code

  10. Borated coolant decontamination and water-chemical regime of the primary coolant circuit of WWER-1 type reactor

    International Nuclear Information System (INIS)

    Matskevich, G.V.; Klement'eva, E.M.; Plotnikov, I.M.

    1974-01-01

    Treatment of the boron concentrate from the evaporation plant by coagulation and mechanical filtration and H-cation exchange, Bo 3 3- anion exchange makes it possible to return the boron concentrate to the primary loop. The duration of the filter cycle, during the course of which the boron concentrate is very effectively treated with respect to all the 'standard' characteristics, is determined by the start of breakthrough of potassium and ammonium cations from the H-cation exchange filter. After regenerating the cation filter with acid the effectiveness of treatment by the anion exchange filter is restored. The investigation leads to the recommendation of a technology for regenerating boric acid not only from the boron-containing drain waters from the primary loop under conditions of liquid boron control but also, if necessary, from the non-design leakage flows to the special drain system of the station. The optimal water conditions for the primary loop of the WWER-1 reactor when effecting control by a boron-containing coolant are those in which the water is treated with caustic potash and a volatile alkali, ammonia or hydrazine hydrate. When working with a coolant which has a high concentration of boric acid (up to 7-8g/l), corrosion-free conditions for the materials of the high- and low-temperature parts of the primary loop can be achieved only by an additional dose of hydrogen-containing alkali

  11. Directional depletion interactions in shaped particles

    Directory of Open Access Journals (Sweden)

    A. Scala

    2014-09-01

    Full Text Available Entropic forces in colloidal suspensions and in polymer-colloid systems are of long-standing and continuing interest. Experiments show how entropic forces can be used to control the self-assembly of colloidal particles. Significant advances in colloidal synthesis made in the past two decades have enabled the preparation of high quality nano-particles with well-controlled sizes, shapes, and compositions, indicating that such particles can be utilized as "artificial atoms" to build new materials. To elucidate the effects of the shape of particles upon the magnitude of entropic interaction, we analyse the entropic interactions of two cut-spheres. We show that the solvent induces a strong directional depletion attraction among flat faces of the cut-spheres. Such an effect highlights the possibility of using the shape of particles to control directionality and strength of interaction.

  12. Zero waste machine coolant management strategy at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Carlson, B.; Algarra, F.; Wilburn, D.

    1998-01-01

    Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extended coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98

  13. Comparative Analysis of VERA Depletion Problems

    International Nuclear Information System (INIS)

    Park, Jinsu; Kim, Wonkyeong; Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung

    2016-01-01

    Each code has its own solver for depletion, which can produce different depletion calculation results. In order to produce reference solutions for depletion calculation comparison, sensitivity studies should be preceded for each depletion solver. The sensitivity tests for burnup interval, number of depletion zones, and recoverable energy per fission (Q-value) were performed in this paper. For the comparison of depletion calculation results, usually the multiplication factors are compared as a function of burnup. In this study, new comparison methods have been introduced by using the number density of isotope or element, and a cumulative flux instead of burnup. In this paper, optimum depletion calculation options are determined through the sensitivity study of the burnup intervals and the number of depletion intrazones. Because the depletion using CRAM solver performs well for large burnup intervals, smaller number of burnup steps can be used to produce converged solutions. It was noted that the depletion intra-zone sensitivity is only pin-type dependent. The 1 and 10 depletion intra-zones for the normal UO2 pin and gadolinia rod, respectively, are required to obtain the reference solutions. When the optimized depletion calculation options are used, the differences of Q-values are found to be a main cause of the differences of solutions. In this paper, new comparison methods were introduced for consistent code-to-code comparisons even when different kappa libraries were used in the depletion calculations

  14. Comparative Analysis of VERA Depletion Problems

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Kim, Wonkyeong; Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-10-15

    Each code has its own solver for depletion, which can produce different depletion calculation results. In order to produce reference solutions for depletion calculation comparison, sensitivity studies should be preceded for each depletion solver. The sensitivity tests for burnup interval, number of depletion zones, and recoverable energy per fission (Q-value) were performed in this paper. For the comparison of depletion calculation results, usually the multiplication factors are compared as a function of burnup. In this study, new comparison methods have been introduced by using the number density of isotope or element, and a cumulative flux instead of burnup. In this paper, optimum depletion calculation options are determined through the sensitivity study of the burnup intervals and the number of depletion intrazones. Because the depletion using CRAM solver performs well for large burnup intervals, smaller number of burnup steps can be used to produce converged solutions. It was noted that the depletion intra-zone sensitivity is only pin-type dependent. The 1 and 10 depletion intra-zones for the normal UO2 pin and gadolinia rod, respectively, are required to obtain the reference solutions. When the optimized depletion calculation options are used, the differences of Q-values are found to be a main cause of the differences of solutions. In this paper, new comparison methods were introduced for consistent code-to-code comparisons even when different kappa libraries were used in the depletion calculations.

  15. A numerical analysis of molten metal drop and coolant interaction

    International Nuclear Information System (INIS)

    Cao, X.; Hajima, Ryoichi; Furuta, Kazuo; Kondo, Shunsuke

    2000-01-01

    The interaction of molten metal drop and coolant is numerically analyzed to investigate the mechanism of fragmentation in vapor explosion. The numerical study is carried out by using a developed simulation code based on multi-phase thermal hydraulic model, which includes physical phenomena required for the analysis: heat transfer, mass change, liquid evaporation and treatment of surface. Several computational techniques are also implemented to improve the efficiency and stability of the numerical scheme. The obtained numerical results show that the growth of spikes on the molten metal drop surface is similar to that observed in Ciccarelli's experiment. The numerical study suggests that quick growth of spikes is the essential mechanism of fragmentation, which is caused by Taylor instability. (author)

  16. Environmental radiological consequences of a loss of coolant accident

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1981-01-01

    The elaboration of a calculation model to determine safety areas, named Exclusion Zone and Low Population Zone for nuclear power plants, is dealt with. These areas are determined from a radioactive doses calculation for the population living around the NPP after occurence of a postulated ' Maximum Credible Accident' (MCA). The MCA is defined as an accident with complete loss of primary coolant and consequent fusion of a substantial portion of the reactor core. In the calculations carried out, data from NPP Angra I were used and the assumptions made were conservative, to be compatible with licensing requirements. Under the most pessimistic assumption (no filters) the values of 410m and 1000m were obtained for the Exclusion Zone and Low Population Zone radii, respectivily. (Author) [pt

  17. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  18. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  19. Reactor coolant pump shaft seal behavior during station blackout

    International Nuclear Information System (INIS)

    Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

    1985-04-01

    A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

  20. The binary cycle for dry-coolant electric power plants

    International Nuclear Information System (INIS)

    Jaumotte, A.L.; Gaivao, A.

    1983-01-01

    Steadily growing water requirements have led the designers of large nuclear power plants to consider the use of a dry coolant seriously in the future (since the problem of a water shortage has already emerged). Two fluids are selected, ammonia and freon 22. In fact, this represents a binary cycle, either with ammonia or with freon (phase change heat transfer fluid). Freon 22 offers the major advantage over ammonia of great chemical stability. It is also nonflammable and nontoxic. Moreover, the use of a phase change fluid offers the possibility of reducing the difference between the air temperature and the steam condensation temperature, since exchange in the condenser is isothermal. The feasibility of the mechanical components, and the successful progress of experiments under way, tend to show that the binary cycle may be a highly attractive answer to the problem of water availability [fr

  1. Assessment of TRIGA pulsing reactor safety without loss of coolant

    International Nuclear Information System (INIS)

    Baers, B.; Kaall, L.

    1979-11-01

    The fuel temperature, fuel can pressure and cladding stresses in the TRIGA fuel elements at various transient states has in part been measured and semiempirically estimated in cases where no experiments could be carried out. To the end of obtaining more reliable and up-to-date data and reactor parameters, reactor power and fuel temperatures were measured both at steady and transient states, including pulsing up to 280 MW. A novel and more realistic model for the fuel temperature at pulsing was also presented and used in the assessments. Based on the present assessment with the water coolant present and a maximum excess reactivity of (Δksub(max)/β = ) 4 $, no reactivity induced transients can bring about any undue hazards to the reactor or to the surroundings. (author)

  2. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  3. A new inexpensive electrochemical meter for oxygen in sodium coolant

    International Nuclear Information System (INIS)

    Periaswami, G.; Rajan Babu, S.S.; Mathews, C.K.

    1987-01-01

    This report describes the development of an inexpensive oxygen meter for sodium coolant and gives the results of the test experiments. Calcia stabilized zirconia has been found to have necessary domain boundary characteristics at low temperatures for use as oxygen sensor in liquid sodium system. It is possible to obtain acceptable sensor cell resistance at temperatures as low as 230 C if K, K 2 O or Na, Na 2 O is used as reference electrode. The performance of these cells has been tested in bench top sodium loops over long periods. Their performance in terms of cell-out put variation with change in oxygen concentration in sodium has been found to be satisfactory. They also have sufficiently long life times since the kinetics of sodium attack on the electrolyte is slow at low temperatures. (author). 17 refs., 6 figs

  4. Insight into steam explosion in stratified melt-coolant configuration

    OpenAIRE

    Grishchenko, Dmitry; Konovalenko, Alexander; Karbojian, Aram; Kudinova, Valtyna; Bechta, Sevostian; Kudinov, Pavel

    2013-01-01

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident  mitigation  strategy.  When  vessel  breach  is  large  and  water  pool  is shallow,  released  corium  melt  can  reach  containment  floor  in  liquid  form  and spread under water creating a stratified configuration of melt covered by coolant. Steam  explosion  in  such  stratified  configuration  was  long  believed  ...

  5. Condensing heat transfer following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Krotiuk, W.J.; Rubin, M.B.

    1978-01-01

    A new method for calculating the steam mass condensation energy removal rates on cold surfaces in contact with an air-steam mixture has been developed. This method is based on the principles of mass diffusion of steam from an area of high concentration to the condensing surface, which is an area of low steam concentration. This new method of calculating mass condensation has been programmed into the CONTEMPT-LT Mod 26 computer code, which calculates the pressure and temperature transients inside a light water reactor containment following a loss-of-coolant accident. The condensing heat transfer coefficient predicted by the mass diffusion method is compared to existing semi-empirical correlations and to the experimental results of the Carolinas Virginia Tube Reactor Containment natural decay test. Closer agreement with test results is shown in the calculation of containment pressure, temperature, and heat sink surface temperature using the mass diffusion condensation method than when using any existing semi-empirical correlation

  6. Experimental Investigation of Heat Transfer Characteristics of Automobile Radiator using TiO2-Nanofluid Coolant

    Science.gov (United States)

    Salamon, V.; Senthil kumar, D.; Thirumalini, S.

    2017-08-01

    The use of nanoparticle dispersed coolants in automobile radiators improves the heat transfer rate and facilitates overall reduction in size of the radiators. In this study, the heat transfer characteristics of water/propylene glycol based TiO2 nanofluid was analyzed experimentally and compared with pure water and water/propylene glycol mixture. Two different concentrations of nanofluids were prepared by adding 0.1 vol. % and 0.3 vol. % of TiO2 nanoparticles into water/propylene glycol mixture (70:30). The experiments were conducted by varying the coolant flow rate between 3 to 6 lit/min for various coolant temperatures (50°C, 60°C, 70°C, and 80°C) to understand the effect of coolant flow rate on heat transfer. The results showed that the Nusselt number of the nanofluid coolant increases with increase in flow rate. At low inlet coolant temperature the water/propylene glycol mixture showed higher heat transfer rate when compared with nanofluid coolant. However at higher operating temperature and higher coolant flow rate, 0.3 vol. % of TiO2 nanofluid enhances the heat transfer rate by 8.5% when compared to base fluids.

  7. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Science.gov (United States)

    2010-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... operated from the control room. (b) The design of the vents and associated controls, instruments and power... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting...

  8. Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

    International Nuclear Information System (INIS)

    Hudcovsky, S.; Slanina, M.; Badiar, S.

    2001-01-01

    The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

  9. Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

    Directory of Open Access Journals (Sweden)

    N. Dharmaraju

    2008-01-01

    Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

  10. Use of Russian technology of ship reactors with lead-bismuth coolant in nuclear power

    International Nuclear Information System (INIS)

    Zrodnikov, A.V.; Chitaykin, V.I.; Gromov, B.F.; Grigoryv, O.G.; Dedoul, A.V.; Toshinsky, G.I.; Dragunov, Yu.G.; Stepanov, V.S.

    2000-01-01

    The experience of using lead-bismuth coolant in Russian nuclear submarine reactors has been presented. The fundamental statements of the concept of using the reactors cooled by lead-bismuth alloy in nuclear power have been substantiated. The results of developments for using lead bismuth coolant in nuclear power have been presented. (author)

  11. Thermo-Hydraulic Behaviour of Coolant in Nuclear Reactor VVER-440 Under Refuelling Conditions

    Directory of Open Access Journals (Sweden)

    Paulech Juraj

    2017-04-01

    Full Text Available The paper presents the numerical simulation of thermo-hydraulic behaviour of coolant in the VVER- 440 nuclear reactor under standard outage conditions. Heating-up and flow of coolant between the reactor pressure vessel and spent fuel storage pool are discussed.

  12. The corrosion products in the coolant circuits of pressurized water nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

  13. Some peculiarities of checking the Topaz-2 system coolant filling quality

    Science.gov (United States)

    Ogloblin, B. G.; Svishchev, A. M.; Shalaev, A. I.

    1996-03-01

    This paper contains the analysis of validity of methods used for checking the Topaz-2 system coolant filling quality by a metering tank according to the mathematical model developed. A number of criteria is proposed for detecting occluded gas in the coolant loop.

  14. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D.P.; Lukas, M.; Lynch, B.K. [Spectro Incorporated, Littleton, MA (United States)

    1997-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  15. Depletion of the Complex Multiple Aquifer System of Jordan

    Science.gov (United States)

    Rödiger, T.; Siebert, C.; Geyer, S.; Merz, R.

    2017-12-01

    In many countries worldwide water scarcity pose a significant risk to the environment and the socio-economy. Particularly in countries where the available water resources are strongly limited by climatic conditions an accurate determination of the available water resources is of high priority, especially when water supply predominantly rely oon groundwater resources and their recharge. If groundwater abstraction exceeds the natural groundwater recharge in heavily used well field areas, overexploitation or persistent groundwater depletion occurs. This is the case in the Kingdom of Jordan, where a multi-layer aquifer complex forms the eastern subsurface catchment of the Dead Sea basin. Since the begin of the industrial and agricultural development of the country, dramatically falling groundwater levels, the disappearance of springs and saltwater intrusions from deeper aquifers is documented nation-wide. The total water budget is influenced by (i) a high climatic gradient from hyperarid to semiarid and (ii) the intnese anthropogenic abstraction. For this multi-layered aquifer system we developed a methodology to evaluate groundwater depletion by linking a hydrological and a numerical flow model including estimates of groundwater abstraction. Hence, we define groundwater depletion as the rate of groundwater abstraction in excess of natural recharge rate. Restricting our analysis, we calculated a range of groundwater depletion from 0% in the eastern Hamad basin to around 40% in the central part of Jordan and to extreme values of 100% of depletion in the Azraq and Disi basin.

  16. Examining depletion theories under conditions of within-task transfer.

    Science.gov (United States)

    Brewer, Gene A; Lau, Kevin K H; Wingert, Kimberly M; Ball, B Hunter; Blais, Chris

    2017-07-01

    In everyday life, mental fatigue can be detrimental across many domains including driving, learning, and working. Given the importance of understanding and accounting for the deleterious effects of mental fatigue on behavior, a growing body of literature has studied the role of motivational and executive control processes in mental fatigue. In typical laboratory paradigms, participants complete a task that places demand on these self-control processes and are later given a subsequent task. Generally speaking, decrements to subsequent task performance are taken as evidence that the initial task created mental fatigue through the continued engagement of motivational and executive functions. Several models have been developed to account for negative transfer resulting from this "ego depletion." In the current study, we provide a brief literature review, specify current theoretical approaches to ego-depletion, and report an empirical test of current models of depletion. Across 4 experiments we found minimal evidence for executive control depletion along with strong evidence for motivation mediated ego depletion. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  17. THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

    International Nuclear Information System (INIS)

    Muramatsu, Ken; Akimoto, Masayuki

    1982-08-01

    THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

  18. Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

    2014-01-01

    The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

  19. The management of the Primary Coolant in HANARO (1998-2006)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Wook; Lee, Mun; Park, Se Il; Hur, Soon Ock; Lim, Kyeong Hwan; Ahn, Guk Hoon

    2007-11-15

    The primary coolant quality management in HANARO is important to prevent corrosion of nuclear fuel and reactor structure material. The purity of the primary coolant has monitored by measuring the electrical conductivity of the primary purification system and pH periodically. The status of the primary coolant quality management was investigated from 1998 to 2006 with the measured data. The electrical conductivity of the primary coolant becomes low as much as the flow-rate of the primary purification system increased because the performance of the resin is in proportion to flow rate through resin. In general, the pH and the electrical conductivity of the primary coolant have been managed within 1 {mu}S/cm which is a target criteria of HANARO.

  20. Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

    International Nuclear Information System (INIS)

    Ruger, C.J.; Luckas, W.J. Jr.

    1989-03-01

    Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

  1. TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

    International Nuclear Information System (INIS)

    Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

    1981-02-01

    This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

  2. Materials, coolants, and components for perspective transmutation power units of VVER spent fuel

    International Nuclear Information System (INIS)

    Matal, O.; Hosnedl, P.

    2004-01-01

    The aim of the contribution is provide information on the selection and development of materials and molten salt coolants for high temperature applications for the new nuclear power units and discuss briefly design problems of selected components such as reactor and salt-salt and salt-gas heat exchangers. The following concepts and topics are discussed : Liquid metal reactor concepts (liquid sodium coolant concepts, lead-bismuth coolant concepts, lead coolant concept), CO 2 -cooled reactor concepts, molten-salt fueled and cooled reactor concepts, existing and prospective molten-salt fuels and coolants and structural materials, structural material development for MSR concepts, experimental studies on material corrosion in fluoride molten-salts, and R and D of some components for MSR. (P.A.)

  3. Uranium, depleted uranium, biological effects

    International Nuclear Information System (INIS)

    2001-01-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  4. Solar receiver protection means and method for loss of coolant flow

    Science.gov (United States)

    Glasgow, L.E.

    1980-11-24

    An apparatus and method are disclosed for preventing a solar receiver utilizing a flowing coolant liquid for removing heat energy therefrom from overheating after a loss of coolant flow. Solar energy is directed to the solar receiver by a plurality of reflectors which rotate so that they direct solar energy to the receiver as the earth rotates. The apparatus disclosed includes a first storage tank for containing a first predetermined volume of the coolant and a first predetermined volume of gas at a first predetermined pressure. The first storage tank includes an inlet and outlet through which the coolant can enter and exit. The apparatus also includes a second storage tank for containing a second predetermined volume of the coolant and a second predetermined volume of the gas at a second predetermined pressure, the second storage tank having an inlet through which the coolant can enter. The first and second storage tanks are in fluid communication with each other through the solar receiver. The first and second predetermined coolant volumes, the first and second gas volumes, and the first and second predetermined pressures are chosen so that a predetermined volume of the coolant liquid at a predetermined rate profile will flow from the first storage tank through the solar receiver and into the second storage tank. Thus, in the event of a power failure so that coolant flow ceases and the solar reflectors stop rotating, a flow rate maintained by the pressure differential between the first and second storage tanks will be sufficient to maintain the coolant in the receiver below a predetermined upper temperature until the solar reflectors become defocused with respect to the solar receiver due to the earth's rotation.

  5. Investigation of loss of coolant accidents in pressurized water reactors using the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method for considering of uncertainties in TRACE

    International Nuclear Information System (INIS)

    Sporn, Michael; Hurtado, Antonio

    2016-01-01

    Loss of coolant accident must take uncertainties with potentially strong effects on the accident sequence prediction into account. For example, uncertainties in computational model input parameters resulting from varying geometry and material data due to manufacturing tolerances or unavailable measurements should be considered. The uncertainties of physical models used by the software program are also significant. In this paper, use of the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method to quantify the uncertainties in the TRACE thermal-hydraulic program is demonstrated. For demonstration purposes loss of coolant accidents with breaks of various types and sizes in a DN 700 reactor coolant pipe are used as an example Application.

  6. Colloids in PWR primary and secondary coolant. Innovative analytical methods

    International Nuclear Information System (INIS)

    Nowotka, Karsten; Guillodo, Michael; Burchardt, Carsten; Geier, Roland; Lehr, Robert; Stellwag, Bernhard

    2014-01-01

    Transport and deposition of corrosion products in the colloid size range between 1 nm and 1 μm are important for heat transfer performance and corrosion in primary and secondary cooling circuits of LWRs. Direct analysis of the properties of small-sized colloids (< 0.45 μm) is difficult due to the pronounced change of the physicochemical properties of coolant samples in sampling lines. An innovative method, based on a filter cascade and developed in the AREVA Technical Center, named 'Colloid Catcher' (CC, patent pending), permits on-line measurements of the properties of corrosion products in the coolant of LWRs. CC measurements are complementary to classic trace analysis addressing the soluble content. Low and high temperature (up to 330°C) test sections are available, depending on our customer's needs. The CC contains differential pressure detectors at each of the three consecutive membrane filters which allow for an in-situ characterization without modification of the corrosion products chemical nature due to temperature changes and subsequent exposure to the atmosphere. With this method, a 'Colloid Fingerprint' of the test solution can be obtained, ideal for an assessment of the transport and deposition of corrosion products in laboratory and on-site studies. The on-line data can of course be complemented by post filtration membrane characterization by digestion and/or optical methods. The high temperature CC serves at the same time as a sampling point for grab samples, with good reproducibility thanks to continuous liquid flow. The CC has been designed to be deployable on laboratory or industrial cooling circuits. The CC test sections have been qualified using AREVA Technical Center's test loops. First test results obtained with the LT CC are presented. Laboratory data can be used to back up existing results and data of on-site measurement campaigns at BWR and PWR plants which were determined with a basic version of the LT CC

  7. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  8. Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

    Directory of Open Access Journals (Sweden)

    Zhi Li

    2017-09-01

    Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

  9. Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

    Science.gov (United States)

    Muruganandam, M.; Mukesh Kumar, P. C.

    2017-10-01

    There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

  10. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  11. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  12. Physics study of Canada deuterium uranium lattice with coolant void reactivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI), Daejeon (Korea, Republic of)

    2017-02-15

    This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 x 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  13. Physics Study of Canada Deuterium Uranium Lattice with Coolant Void Reactivity Analysis

    Directory of Open Access Journals (Sweden)

    Jinsu Park

    2017-02-01

    Full Text Available This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700 fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 × 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  14. Air shipment of radioactive materials in depleted uranium shielded packaging

    Energy Technology Data Exchange (ETDEWEB)

    Adaev, V.A.; Kalevich, E.S. [Federal State Unitary Enterprise ' ' State Scientific Centre of Russia Research Inst. of Atomic Reactors' ' , Dimitrovgrad (Russian Federation); Metelkin, Yu.A.; Orlov, V.K. [Federal State Unitary Enterprise Academician Bochvar, All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation)

    2004-07-01

    Under conditions of strong competition at the international radioisotope market not only the quality of the products themselves but the timeliness and punctuality of a delivery time-table and shortening the delivery period as well became more acute. The latter allows reducing losses of expensive products due to radioactive decay. Distinctive features for arranging transport of radioisotope products using depleted uranium shipping packing sets both in respect of regime of flight and IAEA Regulations (1996) are discussed below.

  15. A Collector Geometry Impact on the Coolant Flow Distribution in the Reactor Model Core

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2015-01-01

    Full Text Available In creating the reactor facility for the transport and energy module of a megawatt class the important task is to optimize a coolant flow path, i.e. to provide a moderate flow resistance and uniform distribution of a coolant. A kind of the chosen collector design to supply coolant significantly contributes to hydraulic losses, in particular, the porosity of the inlet lattice which may lead to uneven coolant rate at the inlet, flow pulsations, and hydraulic losses.For the first time in domestic practice the work examines an impact of the inlet lattices geometry on the averaged and pulsating flow both in a hemispherical collector and at the core inlet to the model paths of a reactor gas-cooled coolant, and gives advices on optimization of collector paths of the coolant flow.The paper presents the results of experiments carried out on the gas dynamic model of the coolant paths containing the inlet lattices of different porosity. It offers a numerical simulation of the flow in the two-parameter model using k-ε turbulence model and ANSYS CFX v14.0 software package and demonstrates a compliance of experimental data with numerical results.The obtained results show that the inlet lattice with a porosity of 0.25 allows relative leveling of the coolant flow directly at the core inlet, which for a uniform cross-sectional energy release reduces temperature of fuel elements. The considered options of design solutions allow you to select the inlet lattice structure, and the core, as well, according to the porosity parameter to solve the problem of reducing hydraulic losses in the coolant paths, reducing pulsating components of the flow in the core and length of the initial portion of flow stabilization. References

  16. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  17. Depleted uranium disposal options evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Hertzler, T.J.; Nishimoto, D.D.; Otis, M.D. [Science Applications International Corp., Idaho Falls, ID (United States). Waste Management Technology Div.

    1994-05-01

    The Department of Energy (DOE), Office of Environmental Restoration and Waste Management, has chartered a study to evaluate alternative management strategies for depleted uranium (DU) currently stored throughout the DOE complex. Historically, DU has been maintained as a strategic resource because of uses for DU metal and potential uses for further enrichment or for uranium oxide as breeder reactor blanket fuel. This study has focused on evaluating the disposal options for DU if it were considered a waste. This report is in no way declaring these DU reserves a ``waste,`` but is intended to provide baseline data for comparison with other management options for use of DU. To PICS considered in this report include: Retrievable disposal; permanent disposal; health hazards; radiation toxicity and chemical toxicity.

  18. Depleted uranium disposal options evaluation

    International Nuclear Information System (INIS)

    Hertzler, T.J.; Nishimoto, D.D.; Otis, M.D.

    1994-05-01

    The Department of Energy (DOE), Office of Environmental Restoration and Waste Management, has chartered a study to evaluate alternative management strategies for depleted uranium (DU) currently stored throughout the DOE complex. Historically, DU has been maintained as a strategic resource because of uses for DU metal and potential uses for further enrichment or for uranium oxide as breeder reactor blanket fuel. This study has focused on evaluating the disposal options for DU if it were considered a waste. This report is in no way declaring these DU reserves a ''waste,'' but is intended to provide baseline data for comparison with other management options for use of DU. To PICS considered in this report include: Retrievable disposal; permanent disposal; health hazards; radiation toxicity and chemical toxicity

  19. A review of hydrodynamic instabilities and their relevance to mixing in molten fuel coolant interactions

    International Nuclear Information System (INIS)

    Fletcher, D.F.

    1984-03-01

    A review of the literature on Rayleigh-Taylor, Kelvin-Helmholtz and capillary instability is presented. The concept of Weber breakup is examined and found to involve a combination of the above instabilities. Sample calculations are given which show how these instabilities may contribute to the mixing of melt and coolant in a molten fuel coolant interaction. It is concluded that Rayleigh-Taylor instability is likely to be important as the melt falls into the coolant and that Kelvin-Helmholtz instability is likely to develop when significant vapour velocities occur. (author)

  20. Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2015-07-01

    Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

  1. Lubrication analysis of the journal bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Kim, J. I.; Jang, M. H.

    2000-01-01

    Special type journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel. The MCP bearings are lubricated with water without external lubricating oil supply. Long bearing with vertical grooves is designed with relatively large bearing clearance to accommodate the long shaft. Lubricational analysis method for journal bearing with vertical grooves in the main coolant pump of SMART is proposed, and lubricational characteristics of the bearings are examined in this paper

  2. Lubrication analysis of the thrust bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Hur, H.; Kim, J. I.

    2001-01-01

    Thrust bearing and journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and especially the MCP bearings are lubricated with water without external lubricating oil supply. Because axial load capacity of the thrust bearing can hardly meet requirement to acquire hydrodynamic or fluid film lubrication state, self-lubrication characteristics of silicon graphite meterials would be needed. Lubricational analysis method for thrust bearing for the main coolant pump of SMART is proposed, and lubricational characteristics of the bearing generated by solving the Reynolds equation are examined in this paper

  3. Investigation of the compatibility between ADS target material with coolant

    International Nuclear Information System (INIS)

    Xu Yongli; Long Bin; Zhang Pingzhu

    2002-01-01

    The compatibility tests for the forging tungsten with coolants are performed in sodium at 500, 600 and 700 degree C, and in water at 100 degree C. The results show: a compact W x O y film is formed at the surface of the tungsten, its thickness is about 1 - 2 μm. There are also corrosion product Na x WO y on the surface of the specimens, the amount of Na x WO y dependents on the temperature and the oxygen content in the sodium. After test for more than 400 h, the matrix of tungsten is not attacked further by sodium and oxygen resulting from the protection of the W x O y film, the thickness of W x O y film and the weight loss become a constant; for the corrosion of the tungsten in water, a W x O y film at the specimen surface is formed at the beginning of the test, its thickness is about 0.8 μm; This film is porous and loose, and they peels off after test for more than 100 h. The new oxide film does not form again because of the lack of the oxygen, the weight loss of the specimens is near a constant

  4. Numerical simulation on coolant flow and heat transfer in core

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis

  5. A study on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H{sub 2} gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs.

  6. Evaluation of zinc addition to PWR primary coolant

    International Nuclear Information System (INIS)

    Pathania, R.; Yagnik, S.; Gold, R.E.; Dove, M.; Kolstad, E.

    1995-01-01

    Laboratory studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. Because of the potential benefits of zinc addition in reducing radiation fields and mitigating PWSCC of Alloy 600 a project was initiated to qualify zinc addition to a PWR. The objective of this work was to evaluate the effect of zinc addition on radiation fields, PWSCC of Alloy 600 and fuel cladding corrosion at the Farley-2 PWR. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor prior to zinc addition at Farley-2. This paper provides an overview of the scope of the zinc addition demonstration at Farley-2 and the fuel cladding corrosion tests at Halden. The zinc concentration in the Farley-2 coolant is approximately 40 ppb and that in Halden is 50 ppb. The paper presents initial results from these studies which are still in progress

  7. Reactor water chemistry relevant to coolant-cladding interaction

    International Nuclear Information System (INIS)

    1987-09-01

    The report is a summary of the work performed in a frame of a Coordinated Research Program organized by the IAEA and carried out from 1981 till 1986. It consists of a survey on our knowledge on coolant-cladding interaction: the basic phenomena, the relevant parameters, their control and the modelling techniques implemented for their assessment. Based upon the results of this Coordinated Research Program, the following topics are reviewed on the report: role of water chemistry in reliable operation of nuclear power plants; water chemistry specifications and their control; behaviour of fuel cladding materials; corrosion product behaviour and crud build-up in reactor circuits; modelling of corrosion product behaviour. This report should be of interest to water chemistry supervisors at the power plants, to experts in utility engineering departments, to fuel designers, to R and D institutes active in the field and to the consultants of these organizations. A separate abstract was prepared for each of the 3 papers included in the Annex of this document. Refs, figs, tabs

  8. Fast reactors bulk sodium coolant disposal NOAH process application

    International Nuclear Information System (INIS)

    Magny, E. de; Berte, M.

    1997-01-01

    Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

  9. A study on safety measure of LMR coolant

    International Nuclear Information System (INIS)

    Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H 2 gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs

  10. Statistical analysis of the Ft. Calhoun reactor coolant pump system

    International Nuclear Information System (INIS)

    Heising, Carolyn D.

    1998-01-01

    In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach to plant maintenance and control, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R-charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specifications limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (author)

  11. A study of the loss of coolant accident

    International Nuclear Information System (INIS)

    Lee, Y.W.; Chung, M.K.; Kim, S.H.; Park, J.S.; Lee, C.B.; Kim, S.B.; Won, S.Y.; Cho, Y.R.

    1983-01-01

    The primary objectives of this project are: (1) To review the published information on LOCA/ECCS study (2) To investigate reflood phenomena and to provide necessary information for analytical model development (3) To modyfy and develop a reflood analysis code. To review the published information on LOCA/ECCS, heat transfer phenomena are divided into 4 regions. Heat transfer correlations published in the references are reviewed and classified according to the regions. To investigate reflood phenomena and to provide better modeling of reflood phenomena, experments have been carried out with an electrically heated 3x3 rod bundle. Heat flux and heat transfer coefficients at the hot surface have been determined from the experimental data by HTC program. The influences of the parameters such as flooding rate, coolant subcooling and power generation on the propagation of rewetting front were also investigated. Calculations obtained from REFLUX code were compared with the experimental data to help an understanding of the reflood heat transfer mechanisms, and then some modifications of the code were provided. Improvements in heat transfer correlations of transition and inverted annular film boiling region, and the logic for the selection of heat transfer regime allowed better estimate for rod temperature behavior. (Author)

  12. High-voltage-compatible, fully depleted CCDs

    Energy Technology Data Exchange (ETDEWEB)

    Holland, Stephen E.; Bebek, Chris J.; Dawson, Kyle S.; Emes, JohnE.; Fabricius, Max H.; Fairfield, Jessaym A.; Groom, Don E.; Karcher, A.; Kolbe, William F.; Palaio, Nick P.; Roe, Natalie A.; Wang, Guobin

    2006-05-15

    We describe charge-coupled device (CCD) developmentactivities at the Lawrence Berkeley National Laboratory (LBNL).Back-illuminated CCDs fabricated on 200-300 mu m thick, fully depleted,high-resistivity silicon substrates are produced in partnership with acommercial CCD foundry.The CCDs are fully depleted by the application ofa substrate bias voltage. Spatial resolution considerations requireoperation of thick, fully depleted CCDs at high substrate bias voltages.We have developed CCDs that are compatible with substrate bias voltagesof at least 200V. This improves spatial resolution for a given thickness,and allows for full depletion of thicker CCDs than previously considered.We have demonstrated full depletion of 650-675 mu m thick CCDs, withpotential applications in direct x-ray detection. In this work we discussthe issues related to high-voltage operation of fully depleted CCDs, aswell as experimental results on high-voltage-compatible CCDs.

  13. Experimental Studies on Breakup and Fragmentation Behavior of Molten Tin and Coolant Interaction

    Directory of Open Access Journals (Sweden)

    Yankai Li

    2017-01-01

    Full Text Available Jet breakup and fragmentation behavior significantly affect the likelihood (and ultimate strength of steam explosion, but it is very challenging to assess the potential damage to reactor cavity due to general lack of knowledge regarding jet breakup phenomena. In this study, the METRIC (mechanism study test apparatus for melt-coolant interaction was launched at Shanghai Jiao Tong University to investigate FCI physics. The first five tests on molten tin and water interactions are analyzed in this paper. Significant breakup and fragmentation were observed without considerable pressure pulse, and intense expansion of droplets in local areas was observed at melt temperature higher than 600°C. The chain interactions of expansion all ceased, however, and there was no energetic steam explosion observed. Quantitative analysis on jet breakup length and debris was studied to investigate the effect of the melt temperature, initial diameter of the jet, and so on. Furthermore, the results of tests were compared with current theories. It is found that melt temperature has strong impact on fragmentation that need to be embodied in advanced fragmentation models.

  14. Qualitative infrared spectral analysis of products adsorbed by silica gel from ditolylmethane coolant and their adsorption isotherm

    International Nuclear Information System (INIS)

    Ermakov, V.A.; Benderskaya, O.S.

    1987-01-01

    The IR-spectral analysis has been applied to study the products adsorbed from ditolylmethane first-circuit coolant, as well as from still bottoms after coolant distillation on silicagel of various makes. The qualitative study of desorbate IR-spectra has shown that they refer to the classes of arylaldehydes, diarylketones and carbonic acids. Under actual conditions first-circuit reactor coolant also has a wide set of products of its radiolysis, therefore the spectrum of coolant oxidaton products must be wider. It is noted that adsorption on silica gel, ASK of oxygen-bearing compounds which are present in ditolyl methane coolant has 2 stages

  15. Is gas in the Orion nebula depleted

    International Nuclear Information System (INIS)

    Aiello, S.; Guidi, I.

    1978-01-01

    Depletion of heavy elements has been recognized to be important in the understanding of the chemical composition of the interstellar medium. This problem is also relevant to the study of H II regions. In this paper the gaseous depletion in the physical conditions of the Orion nebula is investigated. The authors reach the conclusion that very probably no depletion of heavy elements, due to sticking on dust grains, took place during the lifetime of the Orion nebula. (Auth.)

  16. Spearfishing to depletion: evidence from temperate reef fishes in Chile.

    Science.gov (United States)

    Godoy, Natalio; Gelcich, L Stefan; Vásquez, Julio A; Castilla, Juan Carlos

    2010-09-01

    Unreliable and data-poor marine fishery landings can lead to a lack of regulatory action in fisheries management. Here we use official Chilean landing reports and non-conventional indicators, such as fishers' perceptions and spearfishing competition results, to provide evidence of reef fishes depletions caused by unregulated spearfishing. Results show that the three largest and most emblematic reef fishes targeted mainly by spearfishers (> 98% of landings) [Graus nigra (vieja negra), Semicossyphus darwini (sheephead or pejeperro), and Medialuna ancietae (acha)] show signs of depletion in terms of abundance and size and that overall the catches of reef fishes have shifted from large carnivore species toward smaller-sized omnivore and herbivore species. Information from two snorkeling speargun world championships (1971 and 2004, Iquique, Chile) and from fishers' perceptions shows the mean size of reef fish to be declining. Although the ecological consequences of reef fish depletion are not fully understood in Chile, evidence of spearfishing depleting temperate reef fishes must be explicitly included in policy debates. This would involve bans or strong restrictions on the use of SCUBA and hookah diving gear for spearfishing, and minimum size limits. It may also involve academic and policy discussions regarding conservation and fisheries management synergies within networks of no-take and territorial user-rights fisheries areas, as a strategy for the sustainable management of temperate and tropical reef fisheries.

  17. [Nutritional depletion in chronic obstructive pulmonary disease].

    Science.gov (United States)

    Chen, Yan; Yao, Wan-zhen

    2004-10-01

    Chronic obstructive pulmonary disease (COPD) is one of the major diseases worldwide. Nutritional depletion is a common problem in COPD patients and also an independant predictor of survival in these patients. Many data are helpful for determining nutritional depletion, including anthropometric measurement, laboratory markers, body composition analysis (fat-free mass and lean mass), and body weight. The mechanism of nutritional depletion in patients with COPD is still uncertain. It may be associated with energy/metabolism imbalance, tissue hypoxia, systemic inflammation, and leptin/orexin disorders. In patients with nutritional depletion, growth hormone and testosterone can be used for nutritional therapy in addition to nutrition supplementation.

  18. Thermal transfer structures coupling electronics card(s) to coolant-cooled structure(s)

    Science.gov (United States)

    David, Milnes P; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Parida, Pritish R; Schmidt, Roger R

    2014-12-16

    Cooling apparatuses and coolant-cooled electronic systems are provided which include thermal transfer structures configured to engage with a spring force one or more electronics cards with docking of the electronics card(s) within a respective socket(s) of the electronic system. A thermal transfer structure of the cooling apparatus includes a thermal spreader having a first thermal conduction surface, and a thermally conductive spring assembly coupled to the conduction surface of the thermal spreader and positioned and configured to reside between and physically couple a first surface of an electronics card to the first surface of the thermal spreader with docking of the electronics card within a socket of the electronic system. The thermal transfer structure is, in one embodiment, metallurgically bonded to a coolant-cooled structure and facilitates transfer of heat from the electronics card to coolant flowing through the coolant-cooled structure.

  19. Heat transfer and fluid flow aspects of fuel--coolant interactions

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1978-09-01

    A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

  20. Confinement barriers for loss of coolant accidents in the SEAFP reactor plant models

    International Nuclear Information System (INIS)

    Blomquist, R.; Ebert, E.; Gay, J.M.; Mazille, F.; Natalizio, A.; Rolandsson, S.; Ross, W.E.; Shen, K.; Sjoeberg, A.

    1995-01-01

    Loss of coolant accidents may mobilise radioactivity and pressurise confinement barriers thereby making a release to the environment possible. The paper defines the radioactivity confinements and presents principal results from the underlying thermal-hydraulic analyses. (orig.)

  1. Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

    International Nuclear Information System (INIS)

    Megahed, M.M.

    1992-01-01

    An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

  2. Application of heat-resistant non invasive acoustic transducers for coolant control in the NPP pipelines

    International Nuclear Information System (INIS)

    Melnikov, V.; Nigmatulin, B.

    1997-01-01

    The use of ultrasonic waves enables remote testing of the coolant flow, detection of solid and gaseous occlusions and measuring of the water velocity and level. Analysis of the acoustic noise makes it possible to detect coolant leaks and diagnose the state and operation of the rotating mechanisms and bearings. Results are given of the research in the development of highly reliable waveguide-type non-invasive acoustic transducers with a long service life. Examples are given of the use of transducers in various fields of nuclear technology: detection of gas in coolant, indication of the coolant level, control of pipe filling and drainage, measurement of liquid film velocity at the pipe inner surface. (M.D.)

  3. Use of a PKZh-902 instrument for monitoring solid phases in an organic coolant

    International Nuclear Information System (INIS)

    Gavrillin, A.I.; Gagarin, S.I.; Sokolov, V.E.; Zabelin, A.I.

    1986-01-01

    Preliminary tests have been performed with a PKZh-902 in checking feed and circulating coolant, and also in evaluating the performance of the cleaning devices. The test program involved determining the stability of the esnsor materials in ditolymethane, examining the effects of radioactive products and those of the optical characteristics of impurities in the first-loop coolant on the readings and errors of measurement, and checking the scope for using the instrument for continuous monitoring of particle concentrations. Results confirm that the PKZh-902 enables one to monitor the composition and concentration of the solid dispersed phase reliably and with adequate accuracy in the feed coolant. The use for monitoring the loop coolant requires additonal research

  4. Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

    International Nuclear Information System (INIS)

    Luo Bangqi

    2012-01-01

    The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

  5. A study on removal of cobalt from the primary coolant by continuous electrode-ionization with various conducting spacers

    Energy Technology Data Exchange (ETDEWEB)

    Yeon, K.H.; Song, J.H.; Moon, S.H. [Department of Environmental Science and Engineering, Kwangju Inst. of Science and Technology (K-JIST) (Korea, Republic of)

    2002-07-01

    CEDI is a hybrid separation system of electrodialysis and ion exchange processes. This system does not require chemicals to regenerate the ion exchange resin and to concentrate the wastewater. In a CEDI system, the ion exchange resin bed plays a major role in the reduction of the high electrical resistance in the dilute compartment, while the ion exchange membranes lead to depletion and concentration of the solutions in the dilute compartment and concentrate compartment, respectively. The production of high purity water in the primary coolant of a nuclear power plant was investigated using a CEDI process along with various ion-conducting spacers, such as an ion exchange resin (IX), polyurethane-coated ion exchange beads (IEPU), and an ion exchange textile (IET). The ion exchange resin was introduced into the ion-depleting compartments of an electrodialysis (ED) stack, and has been used to reduce the electrical resistance of the stack since ED cannot be applied economically to the treatment of dilute solutions due to their high electrical resistances and the development of the polarization phenomena. However, packing the resin beads in the compartment and assembling the stack is laborious work, while attaining a free flowing solution is difficult because the resin beads are driven downward by gravity in the diluted compartment. Nevertheless, a resin-packed ED stack has recently been developed by Millipore, and is now commercially available from U.S. Filter as industrial units. We set out to prepare improved ion-conducting materials suitable for use in CEDI stacks. To this end, IEPU was prepared using a blending method that produces mixtures of resin beads and powder by allophanate/biuret cross-linking. IET was prepared by the radiation grafting of styrene-fulfonic acid or trimethyl-ammonium chloride onto polypropylene non-woven fabric. (authors)

  6. A study on removal of cobalt from the primary coolant by continuous electrode-ionization with various conducting spacers

    International Nuclear Information System (INIS)

    Yeon, K.H.; Song, J.H.; Moon, S.H.

    2002-01-01

    CEDI is a hybrid separation system of electrodialysis and ion exchange processes. This system does not require chemicals to regenerate the ion exchange resin and to concentrate the wastewater. In a CEDI system, the ion exchange resin bed plays a major role in the reduction of the high electrical resistance in the dilute compartment, while the ion exchange membranes lead to depletion and concentration of the solutions in the dilute compartment and concentrate compartment, respectively. The production of high purity water in the primary coolant of a nuclear power plant was investigated using a CEDI process along with various ion-conducting spacers, such as an ion exchange resin (IX), polyurethane-coated ion exchange beads (IEPU), and an ion exchange textile (IET). The ion exchange resin was introduced into the ion-depleting compartments of an electrodialysis (ED) stack, and has been used to reduce the electrical resistance of the stack since ED cannot be applied economically to the treatment of dilute solutions due to their high electrical resistances and the development of the polarization phenomena. However, packing the resin beads in the compartment and assembling the stack is laborious work, while attaining a free flowing solution is difficult because the resin beads are driven downward by gravity in the diluted compartment. Nevertheless, a resin-packed ED stack has recently been developed by Millipore, and is now commercially available from U.S. Filter as industrial units. We set out to prepare improved ion-conducting materials suitable for use in CEDI stacks. To this end, IEPU was prepared using a blending method that produces mixtures of resin beads and powder by allophanate/biuret cross-linking. IET was prepared by the radiation grafting of styrene-fulfonic acid or trimethyl-ammonium chloride onto polypropylene non-woven fabric. (authors)

  7. Theory structure on 'decontamination technology for primary coolant loop of in-service pressurized water reactors'

    International Nuclear Information System (INIS)

    Zhou Xianyu

    1996-07-01

    Theory structure on 'Decontamination Technology for Primary Coolant Loop of In-Service Pressurized Water Reactors' is proposed. The paper aims at systematizing the experiences accumulated abroad in the development of decontamination over a long period of time and raising them to the level of instructive theory, so as to help form 'Decontamination Technology for Primary Coolant Loop of In-Service Pressurized Water Reactors', turn 'skill' into 'science', and stop acting blindly in the development of decontamination. (26 refs., 5 figs.)

  8. Research on model of additional forces of ocean conditions in one-dimensional coolant channel

    International Nuclear Information System (INIS)

    Qian Libo; Tian Wenxi; Qiu Suizheng; Su Guanghui; Li Yong; Huang Yanping; Yan Xiao

    2012-01-01

    The effect of different ocean conditions on coolant flow can come down to the differences of additional forces in the momentum equations, thus ocean conditions can be considered by adding the additional forces caused by them to the momentum equations. The model of additional forces of 6 types of typical and relevant coupled ocean conditions is obtained based on the basic momentum equation in the non-inertial reference frame and the one-dimensional coolant channel. (authors)

  9. Corrosion particles in the primary coolant of VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vajda, N.; Molnar, Z., E-mail: vajdanor@gmail.com [RadAnal Ltd., Budapest (Hungary); Macsik, Z.; Szeles, E.; Hargittai, P., E-mail: szeles@iki.kfki.hu [Inst. of Isotopes of the Hungarian Academy of Sciences, Budapest (Hungary); Csordas, A.; Pinter, T., E-mail: csordas@aeki.kfki.hu [Atomic Energy Research Inst. of the Hungarian Academy of Sciences, Budapest (Hungary); Pinter, T., E-mail: pinter@npp.hu [Paks Nuclear Power Plant, Paks (Hungary)

    2010-07-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. {sup 60}Co, {sup 54}Mn, {sup 63}Ni, {sup 55}Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of {sup 60}Co/Co, {sup 54}Mn/Fe, {sup 55}Fe/Fe and {sup 63}Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a

  10. Corrosion particles in the primary coolant of VVER-440 reactors

    International Nuclear Information System (INIS)

    Vajda, N.; Molnar, Z.; Macsik, Z.; Szeles, E.; Hargittai, P.; Csordas, A.; Pinter, T.; Pinter, T.

    2010-01-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. 60 Co, 54 Mn, 63 Ni, 55 Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of 60 Co/Co, 54 Mn/Fe, 55 Fe/Fe and 63 Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a result of corrosion of the steel

  11. Plutonium in depleted uranium penetrators

    International Nuclear Information System (INIS)

    McLaughlin, J.P.; Leon-Vintro, L.; Smith, K.; Mitchell, P.I.; Zunic, Z.S.

    2002-01-01

    Depleted Uranium (DU) penetrators used in the recent Balkan conflicts have been found to be contaminated with trace amounts of transuranic materials such as plutonium. This contamination is usually a consequence of DU fabrication being carried out in facilities also using uranium recycled from spent military and civilian nuclear reactor fuel. Specific activities of 239+240 Plutonium generally in the range 1 to 12 Bq/kg have been found to be present in DU penetrators recovered from the attack sites of the 1999 NATO bombardment of Kosovo. A DU penetrator recovered from a May 1999 attack site at Bratoselce in southern Serbia and analysed by University College Dublin was found to contain 43.7 +/- 1.9 Bq/kg of 239+240 Plutonium. This analysis is described. An account is also given of the general population radiation dose implications arising from both the DU itself and from the presence of plutonium in the penetrators. According to current dosimetric models, in all scenarios considered likely ,the dose from the plutonium is estimated to be much smaller than that due to the uranium isotopes present in the penetrators. (author)

  12. Correct numerical simulation of a two-phase coolant

    Science.gov (United States)

    Kroshilin, A. E.; Kroshilin, V. E.

    2016-02-01

    Different models used in calculating flows of a two-phase coolant are analyzed. A system of differential equations describing the flow is presented; the hyperbolicity and stability of stationary solutions of the system is studied. The correctness of the Cauchy problem is considered. The models' ability to describe the following flows is analyzed: stable bubble and gas-droplet flows; stable flow with a level such that the bubble and gas-droplet flows are observed under and above it, respectively; and propagation of a perturbation of the phase concentration for the bubble and gas-droplet media. The solution of the problem about the breakdown of an arbitrary discontinuity has been constructed. Characteristic times of the development of an instability at different parameters of the flow are presented. Conditions at which the instability does not make it possible to perform the calculation are determined. The Riemann invariants for the nonlinear problem under consideration have been constructed. Numerical calculations have been performed for different conditions. The influence of viscosity on the structure of the discontinuity front is studied. Advantages of divergent equations are demonstrated. It is proven that a model used in almost all known investigating thermohydraulic programs, both in Russia and abroad, has significant disadvantages; in particular, it can lead to unstable solutions, which makes it necessary to introduce smoothing mechanisms and a very small step for describing regimes with a level. This does not allow one to use efficient numerical schemes for calculating the flow of two-phase currents. A possible model free from the abovementioned disadvantages is proposed.

  13. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  14. Analysis of Loss-of-Coolant Accidents in the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  15. Seismic response analysis of a reactor coolant pump

    International Nuclear Information System (INIS)

    Villasor, A.P. Jr.

    1975-01-01

    The Westinghouse Type 93A reactor coolant pump (RCP) is analyzed for seismic response. Simply described the RCP is a vertical, single-stage, centrifugal pump designed to move 90,000gpm (568m 3 /sec) of water and driven by a 6000hp motor for use in the PWR primary system. The RCP assembly is generally axisymmetric and is modeled using 3-dimensional elements to represent its parts. These finite elements are of the types normally found in general purpose computer programs such as ANSYS or NASTRAN. The structural frame and the rotating shaft are the principal branches of the model. Each consists of a series of pipe elements complemented by mass elements. Orthogonal sets of linear spring elements connect the branches at the bearings and possibly at each labyrinth. Fluid elements are added to include the interaction between the shaft and the pump case through the intervening water mass. To complete the model, stiffness matrix elements representing the support structure and the neighboring loop piping are attached. It is impractical to idealize faithfully each geometric irregularity. Several adjacent sections are combined into one suitable element with total stiffness and mass equivalence. The number of elements in the model is thus minimized. Shear deflection of the pipe elements is considered; mass and mass inertia are lumped at nodal points, as needed, to compensate for the actual material distribution. The in-depth analysis in this work, believed to have no published precedent, has found the RCP adequate to withstand the imposed seismic loading. All component stresses were within the applicable faulted criteria and the relative movements between closely-mated parts fell inside their nominal clearance limits

  16. Conditioning alkaline coolant radioactive waste from research reactor BR-10

    International Nuclear Information System (INIS)

    Vladimir, Smykov; Mikhail, Kononyuk; Kirill, Butov

    2014-01-01

    In the Institute for Physics and Power Engineering (Russia) has developed and was successfully demonstrated a technology of solid-phase oxidation of alkaline metal by slag from the copper-smelting industry. Neutralization of alkaline metal in the solid-phase oxidation process occurs in a single phase. The solid-phase oxidation process does not result in the generation of hydrogen. The product of alkaline metal radioactive waste processing is solid mineral-like sinter of reaction products, contained inside a steel reaction container, which is immediately shipped for dry storage in a solid radioactive waste storage facility. The presence of a mercury admixture in the research reactor BR-10 (BR-10) reactor alkaline metals radioactive waste makes conditioning of that waste considerably more complicated. Laboratory research demonstrated that mercury could be effectively removed from alkaline metal by pushing the Na-K alloy through chips of metallic magnesium in elevated temperatures. For neutralization of non-drainable sodium residues and admixtures in individual equipment (cold traps, pipe lines, tanks) of the research reactor BR-10 has developed a method for neutralization of non-drainable residues of alkaline liquid metal coolants with a gaseous sub oxide of nitrogen, which is characterized by absence of hydrogen generation, improving the safety of the technology. Currently, the reactor building is undergoing installation of the experimental-industrial plant 'Magma', the purpose of which is processing of accumulated alkaline metals radioactive waste. In according with concept of 'experimental polygon for testing the decommissioning technologies of the BN series of reactors' based on the BR-10 installation, it would appear sensible to start the development of the installation for conditioning by solid-phase oxidation of up to 1000 liters of radioactive waste per loading. (author)

  17. Application of an Underwater Robot in Reactor Coolant System

    International Nuclear Information System (INIS)

    Choi, Young-Soo; Jeong, Kyung-Min; Lee, Sung-Uk; Cho, Jai-Wan

    2006-01-01

    Nuclear energy is a major source of electric energy consumed in Korea. It has the advantage of other energy sources, nuclear energy is cost effective and little pollution. But the fearfulness of an accident and/or failure has scared us the utilization of nuclear energy extensively. So, the safety and reliability of nuclear power plants become more important. Inspection and maintenance of component should be achieved continuously. The RCS(reactor coolant system) of PWR(pressurized water reactor) has a role to cool down the reactor's temperature. Cooling water is injected through the SI(safety injection) nozzle into the cold leg of the primary loop. Thermal sleeves are attached inside the cylindrical SI nozzle to reduce the thermal shock of the cooling water to the weld zone of the safety injection nozzle. The human workers are susceptible to radiation exposure and manual handling machine is hard to access because of the complexity of the path. So, we developed and applied free running, tele-operated underwater vehicle to inspect SI nozzle close to the place. Tele-operated robot is useful to inspect and maintain the component of nuclear power plants to reduce the radiation exposure of human operators and improve the reliability of the operation in nuclear power plants. Underwater robot is comprised of two parts; one is robot vehicle and the other is remote control module. Underwater robot vehicle has 4 DOF(degree of freedom) of mobility and 1 DOF of camera observation. The task to inspect the internal of RCS in nuclear power plant is achieved successfully. And the reliability for the maintenance is increased by the aid of tele-operated robot

  18. Application of an Underwater Robot in Reactor Coolant System

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young-Soo; Jeong, Kyung-Min; Lee, Sung-Uk; Cho, Jai-Wan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    Nuclear energy is a major source of electric energy consumed in Korea. It has the advantage of other energy sources, nuclear energy is cost effective and little pollution. But the fearfulness of an accident and/or failure has scared us the utilization of nuclear energy extensively. So, the safety and reliability of nuclear power plants become more important. Inspection and maintenance of component should be achieved continuously. The RCS(reactor coolant system) of PWR(pressurized water reactor) has a role to cool down the reactor's temperature. Cooling water is injected through the SI(safety injection) nozzle into the cold leg of the primary loop. Thermal sleeves are attached inside the cylindrical SI nozzle to reduce the thermal shock of the cooling water to the weld zone of the safety injection nozzle. The human workers are susceptible to radiation exposure and manual handling machine is hard to access because of the complexity of the path. So, we developed and applied free running, tele-operated underwater vehicle to inspect SI nozzle close to the place. Tele-operated robot is useful to inspect and maintain the component of nuclear power plants to reduce the radiation exposure of human operators and improve the reliability of the operation in nuclear power plants. Underwater robot is comprised of two parts; one is robot vehicle and the other is remote control module. Underwater robot vehicle has 4 DOF(degree of freedom) of mobility and 1 DOF of camera observation. The task to inspect the internal of RCS in nuclear power plant is achieved successfully. And the reliability for the maintenance is increased by the aid of tele-operated robot.

  19. Reduction of radioactivity in the reactor coolant system

    International Nuclear Information System (INIS)

    Takashima, Yoshie; Oosumi, Katsumi; Uchida, Shunsuke; Izumiya, Masakiyo

    1980-01-01

    In evaluating the radioactivity in a reactor coolant system so far, the fission products leaking from fuel rods and the corrosion products activated in a core have been the objects to be considered. However, the target of dose rate reduction is being concentrated only to the suppression of the generation of activated corrosion products, as the former problem was solved by the improvement of operating techniques and fuel manufacturing workmanship. In examining BWR cooling systems, the sources of generating corrosion products are mainly the component materials of feed water and condensate systems and the internal structures of a reactor. Of these, the contribution to dose rate by the former is much larger. Accordingly the relationship between the concentration of corrosion products in feed water and the piping surface dose rate in recirculation system has been analyzed for Shimane and Tsuruga nuclear power stations, investigating the operation results. This analysis has clarified that the insoluble iron cruds and cobalt ions carried into the reactors from the feed water systems were playing an important role for raising the dose rate. The investigation into the suppression of these matters has been done as the main countermeasure for the reduction of radioactivity. As a result, it is expected that the dose rate can be suppressed to 100 mR/h or less in almost all plants if the water quality control is practiced so as to maintain the iron cruds in feed water to the level of 1 ppb or so. The oxygen injection to feed water systems has also large effect on the improvement of iron crud rejection. (Wakatsuki, Y.)

  20. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  1. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, D.D. [Center for Electrochemical Science and Technology, The Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O{sub 2}, H{sub 2}, H{sub 2}O{sub 2}), external solution composition (concentrations of O{sub 2}, H{sub 2}O{sub 2}, and H{sub 2}), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

  2. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

  3. Analysis of a water-coolant leak into a very high-temperature vitrification chamber

    International Nuclear Information System (INIS)

    Felicione, F. S.

    1998-01-01

    A coolant-leakage incident occurred during non-radioactive operation of the Plasma Hearth Process waste-vitrification development system at Argonne National Laboratory when a stray electric arc ruptured az water-cooling jacket. Rapid evaporation of the coolant that entered the very high-temperature chamber pressurized the normally sub-atmospheric system above ambient pressure for over 13 minutes. Any positive pressurization, and particularly a lengthy one, is a safety concern since this can cause leakage of contaminants from the system. A model of the thermal phenomena that describe coolant/hot-material interactions was developed to better understand the characteristics of this type of incident. The model is described and results for a variety of hypothetical coolant-leak incidents are presented. It is shown that coolant leak rates above a certain threshold will cause coolant to accumulate in the chamber, and evaporation from this pool can maintain positive pressure in the system long after the leak has been stopped. Application of the model resulted in reasonably good agreement with the duration of the pressure measured during the incident. A closed-form analytic solution is shown to be applicable to the initial leak period in which the peak pressures are generated, and is presented and discussed

  4. The Chemistry and Toxicology of Depleted Uranium

    Directory of Open Access Journals (Sweden)

    Sidney A. Katz

    2014-03-01

    Full Text Available Natural uranium is comprised of three radioactive isotopes: 238U, 235U, and 234U. Depleted uranium (DU is a byproduct of the processes for the enrichment of the naturally occurring 235U isotope. The world wide stock pile contains some 1½ million tons of depleted uranium. Some of it has been used to dilute weapons grade uranium (~90% 235U down to reactor grade uranium (~5% 235U, and some of it has been used for heavy tank armor and for the fabrication of armor-piercing bullets and missiles. Such weapons were used by the military in the Persian Gulf, the Balkans and elsewhere. The testing of depleted uranium weapons and their use in combat has resulted in environmental contamination and human exposure. Although the chemical and the toxicological behaviors of depleted uranium are essentially the same as those of natural uranium, the respective chemical forms and isotopic compositions in which they usually occur are different. The chemical and radiological toxicity of depleted uranium can injure biological systems. Normal functioning of the kidney, liver, lung, and heart can be adversely affected by depleted uranium intoxication. The focus of this review is on the chemical and toxicological properties of depleted and natural uranium and some of the possible consequences from long term, low dose exposure to depleted uranium in the environment.

  5. The sensitivity of polar ozone depletion to proposed geoengineering schemes.

    Science.gov (United States)

    Tilmes, Simone; Müller, Rolf; Salawitch, Ross

    2008-05-30

    The large burden of sulfate aerosols injected into the stratosphere by the eruption of Mount Pinatubo in 1991 cooled Earth and enhanced the destruction of polar ozone in the subsequent few years. The continuous injection of sulfur into the stratosphere has been suggested as a "geoengineering" scheme to counteract global warming. We use an empirical relationship between ozone depletion and chlorine activation to estimate how this approach might influence polar ozone. An injection of sulfur large enough to compensate for surface warming caused by the doubling of atmospheric CO2 would strongly increase the extent of Arctic ozone depletion during the present century for cold winters and would cause a considerable delay, between 30 and 70 years, in the expected recovery of the Antarctic ozone hole.

  6. Testing strong interaction theories

    International Nuclear Information System (INIS)

    Ellis, J.

    1979-01-01

    The author discusses possible tests of the current theories of the strong interaction, in particular, quantum chromodynamics. High energy e + e - interactions should provide an excellent means of studying the strong force. (W.D.L.)

  7. Polo-like kinase (Plk)1 depletion induces apoptosis in cancer cells.

    Science.gov (United States)

    Liu, Xiaoqi; Erikson, Raymond L

    2003-05-13

    Elevated expression of mammalian polo-like kinase (Plk)1 occurs in many different types of cancers, and Plk1 has been proposed as a novel diagnostic marker for several tumors. We used the recently developed vector-based small interfering RNA technique to specifically deplete Plk1 in cancer cells. We found that Plk1 depletion dramatically inhibited cell proliferation, decreased viability, and resulted in cell-cycle arrest with 4 N DNA content. The formation of dumbbell-like chromatin structure suggests the inability of these cells to completely separate the sister chromatids at the onset of anaphase. Plk1 depletion induced apoptosis, as indicated by the appearance of subgenomic DNA in fluorescence-activated cell-sorter (FACS) profiles, the activation of caspase 3, and the formation of fragmented nuclei. Plk1-depletion-induced apoptosis was partially reversed by cotransfection of nondegradable mouse Plk1 constructs. In addition, the p53 pathway was shown to be involved in Plk1-depletion-induced apoptosis. DNA damage occurred in Plk1-depleted cells and inhibition of ATM strongly potentiated the lethality of Plk1 depletion. Although p53 is stabilized in Plk1-depleted cells, DNA damage also occurs in p53(-/-) cells. These data support the notion that disruption of Plk1 function could be an important application in cancer therapy.

  8. Ozone depletion, related UVB changes and increased skin cancer incidence

    Science.gov (United States)

    Kane, R. P.

    1998-03-01

    plant life also. However, because the increased cancer incidence observed so far may not be (entirely) due to ozone depletion, other causes need to be investigated urgently and, if possible, remedied. Otherwise, deaths due to skin cancer will continue even after CFC production is controlled and ozone levels are recovered. There is no room for complacency. If nothing else is possible, use of protective screens and creams and avoiding exposure to sunlight during peak hours (10:00-15:00 h) should be strongly recommended.

  9. Children's Use of Metaphors in Relation To Their Mental Models: The Case of the Ozone Layer and Its Depletion.

    Science.gov (United States)

    Christidou, Vasilia; Koulaidis, Vasilis; Christidis, Theodor

    1997-01-01

    Examines the relationship between children's use of metaphors and their mental models concerning the ozone layer and ozone layer depletion. Results indicate that the way children represent the role and depletion of ozone is strongly correlated with the types of metaphors they use while constructing and/or articulating their models. Also discusses…

  10. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  11. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei; Sun, Xiaokai [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Jiang, Peixue, E-mail: jiangpx@tsinghua.edu.cn [Key Laboratory for Thermal Science and Power Engineering of Ministry of Educations, Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China); Wang, Jie [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  12. Numerical study on seismic response of the reactor coolant pump in Advanced Passive Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De, Cheng, E-mail: 0100209064@sjtu.edu.cn; Zhen-Qiang, Yao, E-mail: zqyaosjtu@gmail.com; Ya-bo, Xue; Hong, Shen

    2014-10-15

    Highlights: • An artificial accelerogram of the specified SSE is generated. • A dynamic FE model of the RCP in AP1000 (with gyroscopic and FSI effects) is developed. • The displacement, force, moment and stress in the RCP during the earthquake are summarized. - Abstract: The reactor coolant pump in the Advanced Passive Pressurized Water Reactor is a kind of nuclear canned-motor pump. The pump is classified as Seismic Category I, which must function normally during the Safe Shutdown Earthquake. When the nuclear power plant is located in seismically active region, the seismic response of the reactor coolant pump may become very important for the safety assessment of the whole nuclear power plant. In this article, an artificial accelerogram is generated. The response spectrum of the artificial accelerogram fits well with the design acceleration spectrum of the Safe Shutdown Earthquake. By applying the finite element modeling method, the dynamic finite element models of the rotor and stator in the reactor coolant pump are created separately. The rotor and stator are coupled by the journal bearings and the annular flow between the rotor and stator. Then the whole dynamic model of the reactor coolant pump is developed. Time domain analysis which uses the improved state-space Newmark method of a direct time integration scheme is carried out to investigate the response of the reactor coolant pump under the horizontal seismic load. The results show that the reactor coolant pump responds differently in the direction of the seismic load and in the perpendicular direction. During the Safe Shutdown Earthquake, the displacement response, the shear force, the moment and the journal bearing reaction forces in the reactor coolant pump are analyzed.

  13. Fully Depleted Charge-Coupled Devices

    Energy Technology Data Exchange (ETDEWEB)

    Holland, Stephen E.

    2006-05-15

    We have developed fully depleted, back-illuminated CCDs thatbuild upon earlier research and development efforts directed towardstechnology development of silicon-strip detectors used inhigh-energy-physics experiments. The CCDs are fabricated on the same typeof high-resistivity, float-zone-refined silicon that is used for stripdetectors. The use of high-resistivity substrates allows for thickdepletion regions, on the order of 200-300 um, with corresponding highdetection efficiency for near-infrared andsoft x-ray photons. We comparethe fully depleted CCD to thep-i-n diode upon which it is based, anddescribe the use of fully depleted CCDs in astronomical and x-ray imagingapplications.

  14. Novel electromagnetic technique for repositioning of coolant tube spacers in CANDU nuclear reactors

    Science.gov (United States)

    Dableh, Joseph H.

    1986-06-01

    A novel electromagnetic technique to reposition the coolant tube spacers in the fuel channels of CANDU nuclear reactors was successfully developed in the fall of 1983 at Ontario Hydro Research Division. The need to reposition dislocated spacers in noncommissioned reactors was discovered subsequent to the rupture of a pressure tube in one reactor at the Pickering Nuclear Generator Station in Ontario. A contributing factor to the failure of the tube was the fact that the annular spacers (garter springs), used to maintain the coaxial configuration between the pressure tube and its surrounding calandria tube, had been displaced longitudinally for a number of years. Subsequent to this finding, it was discovered that a number of garter springs in noncommissioned nuclear reactors were displaced due to vibration induced by various sources during the construction stage. Since the garter springs are not directly accessible by mechanical means, extensive dismantling of the fuel channels would have been necessary to reposition the springs in their designated locations. This paper describes a novel method to reposition the garter springs without dismantling the fuel channels. The method consists of exerting a force on the springs in the direction of the required displacement by applying a large electromagnetic impulse (generated by a 200-kJ capacitor bank) to a drive coil inserted into the pressure tube opposite the spacer. The repositioning of displaced garter springs in five new reactors in Ontario has been carried out successfully in 1984. The saving in reactor repair cost, interest charges, and replacement energy cost was on the order of hundreds of millions of dollars. Equally large benefits and savings will be realized if the need to use this technique in commissioned reactors arises. Also, the related development of strong compact coils and low-resistance pulse power cable have significant implications and advantages in various other applications related to the pulse

  15. Numerical and experimental hydrodynamic study of a coolant distributor for grinding applications

    Directory of Open Access Journals (Sweden)

    Tala Moussa

    2016-01-01

    Full Text Available In grinding, the high frictional energy is converted into heat, which may cause thermal damage and degradation of the wheel and the workpiece. Unwanted thermal effects must thus be reduced, often by external cooling using a curved-duct coolant distributor to match the wheel geometry. The performance of such a system depends strongly on the impinging jet flow properties to ensure efficient sprinkling of the hot spots. The fluid distributor, placed above the workpiece, is pierced with a certain number of identical nozzle fittings, providing multiple jets at the outlet of the nozzles. These jets sprinkle the solids over a given zone and remove the heat by convective transfer. The cooling is hence dependent on the flow structure, meaning the jet diameters, trajectories and velocities, determined up-flow by the distributor design. The present study is devoted to the hydrodynamics aspects of the fluid distributor, aiming to determine the flow-rate distribution at the different orifices and the flow-rate–pressure relationship, for a variety of nozzle diameters and feeding flow rates, under isothermal conditions. A simple hydraulic balance in the device was not able to predict with sufficient accuracy the actual measurements, even when the Venturi effect was accounted for. This discrepancy is due to the curvature of the distributor, inducing secondary flows in interaction with the nozzle outlets, which leads to a rather complex flow pattern. To overcome this issue, a computational fluid dynamics (CFD tool was used and compared with in situ experiments – global flow rate and pressure measurements were additionally taken with particle image velocimetry (PIV to gain insight into the local structure. Simulations were performed with a 3D turbulence model for Reynolds numbers up to 100,000. This model provides an efficient tool for coupling with the thermal study at a later step, allowing global sizing and energetic optimization of the grinding process.

  16. Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

    International Nuclear Information System (INIS)

    Sharma, V.K.; Surendar, C.; Bapat, C.N.

    1994-01-01

    The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

  17. Modelling nonstationary thermohydrodynamic processes in heat-exchange circuits with a two-phase coolant

    International Nuclear Information System (INIS)

    Blinkov, V.N.

    1993-01-01

    This paper presents a mathematical model and a open-quotes fastclose quotes computer program for analyzing nonstationary thermohydrodynamic processes in distributed multi-element circuits containing a two-phase coolant. The author's approach is based on representing the distributed multi-element circuits with the two-phase coolant (such as cooling circuits of the reactor of an atomic power station) in the form of equivalent thermohydrodynamic chains composed of idealized elements with the intrinsic properties of the structure elements of real systems. The author has developed the nomenclature of such conceptual elements for objects which can be modelled; the nomenclature encompasses the control volumes (with a single-phase or two-phase coolant or a moving boundary of boiling/condensation) and the branch lines (type of tube and connections in dependence on the inertia of the coolant being taken into account) for a hydrodynamic submodel and the thermal components and lines for a thermal submodel. The mathematical models which have been developed and the program using them are designated for various forms of calculating slow thermohydrodynamic processes in multi-element coolant circuits in reactors and modeling test stands. The program facilitates calculation of the range of stable operation, detailed studies of stationary and nonstationary modes of operation, and forecasts of effective engineering measures to obtain stability with the aid of microcomputers

  18. In-core failure of the instrumented BWR rod by locally induced high coolant temperature

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

  19. The determination of magnesium in simulated PWR coolant by graphite furnace atomic absorption spectrometry

    International Nuclear Information System (INIS)

    Gatford, C.; Torrance, K.

    1988-06-01

    The determination of magnesium in simulated PWR coolant has been investigated by graphite furnace atomic absorption spectrometry with atomization from a L'vov platform. The presence of boric acid in the coolant suppresses the magnesium absorption to such an extent that removal of the boron is necessary and three variations of a methyl borate volatilization technique for the in situ removal of boron from the sample platform were investigated. This work has shown that dilution of the sample with an equal volume of acidified methanol and volatilization of the methyl borate was adequate for the determination of magnesium in coolant samples containing up to 2000 mg 1 -1 of boron. In simulated coolant samples containing 25 and 4 μg 1 -1 of magnesium, positive biases of about 2 and 0.5 μg 1 -1 were measured and these errors were considered to be due to contamination. The limit of detection in the presence of 100 and 2000 mg 1 -1 boron were 0.14 and 0.93 μg 1 -1 respectively. These performance characteristics suggest the method is completely acceptable for monitoring the chemical purity of PWR coolant and associated waters containing boric acid. If, however, more precise analyses were to be required for research purposes then any significant improvement in the above figures would require increased purity of reagents, clean-room conditions to reduce contamination and a more versatile atomic absorption spectrophotometer. (author)

  20. Coolant void effect investigation - case of a na-cooled fast reactor

    International Nuclear Information System (INIS)

    Glinatsis, G.; Gugiu, D.

    2013-01-01

    In the frame of the last EURATOM-FP7 Program, a large sized Sodium-cooled FR (SFR) has been studied. Mixed carbides fuel (U, Pu)C has been adopted for the backup core solution and important work has been also performed in order to obtain an ''optimised'' backup configuration ''close'' to the reference one, which is fueled by mixed oxides fuel (U, Pu)Ox. The peculiarity of both core designs (the reference configuration and the optimised backup configuration) is the adoption of a 60 cm Plenum zone in the upper part of each fuel assembly (FA), that is filled by coolant, in order to mitigate (when emptied) the core positive coolant void effect. This paper presents some results of a detailed study of the coolant void effect for the above SFR with mixed carbides core. Many aspects, like geometric heterogeneity, the burnup state, the operating conditions, etc., have been taken into consideration in order to obtain information about the ''propagation'' and the behaviour of the coolant void effect itself. The performed study investigates also the coolant void effect consequences on some reactivity coefficients, which are important for a safe behaviour of the reactor. The investigation consisted in the steady state simulations of the reactor on different operating conditions in Monte Carlo approach. (authors)

  1. Evaluation of primary coolant pH operation methods for the domestic PWRs

    International Nuclear Information System (INIS)

    Paek, Seung Woo; Na, Jung Won; Kim, Yong Eak; Bae, Jae Heum

    1992-01-01

    Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of-core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed. (Author)

  2. Reactor coolant pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  3. Dryout heat flux in a debris bed with forced coolant flow from below

    International Nuclear Information System (INIS)

    Bang, Kwang-Hyun; Kim, Jong-Myung

    2004-01-01

    The objective of the present study is to experimentally investigate the enhancement of dryout heat flux in debris beds with coolant flow from below. The experimental facility consists mainly of an induction heater (40 kW, 35 kHz), a double-wall quartz-tube test section containing steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of particle bed was achieved by induction heating. This paper reports the experimental data for 5 mm particle bed and 300 mm bed height. The dryout heat rate data were obtained of both top-flooding case and forced coolant injection from below with the injection mass flux up to 1.5 kg/m 2 s. For the top-flooded case, the volumetric dryout heat rate was about 4 MW/m 3 and it increased as the rate of coolant injection from below was increased. At the coolant injection mass flux of 1.5 kg/m 2 s, the volumetric dryout heat rate was about 10 MW/m 3 , the enhancement factor was more than two. (author)

  4. Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

    International Nuclear Information System (INIS)

    Bourgeois, P.; Stora, J.P.

    1979-01-01

    The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

  5. Study of heat and mass transfer in a steam generator with chemically reacting coolant

    International Nuclear Information System (INIS)

    Lemeshev, V.U.; Mikhalevich, A.A.; Nemtsev, V.A.; Nesterenko, V.B.

    1983-01-01

    A one-dimensional mathematical model is represented once-through type and heat and mass transfer steam generator with turbulent flow of chemically reacting N 2 O 4 -NO coolant is investigated. During development of the mathematical model it has been assumed that the process of heating and boiling of liquid N 2 O 4 -NO coolant as well as superheating of produced vapour at subcritical parameters or heating of pseudo-liquid and superheating of produced pseudovapour at supercritical parameters (the heated side) is carried out at the expense of gaseous N 2 O 4 -NO coolant cooling (the heating side). The process of heating and cooling of the N 2 O 4 -NO system is followed by N 2 O 4 reversible 2NO 2 (1); 2NO 2 reversible 2NO+O 2 (2); N 2 O 3 reVersible NO 2 +NO (3) reactions, whereas the reactions (1) and (3) are practically equilibrium and the reaction (2) proceeds for the time comparable with the coolant residence time in the reactor circuit and the reaction rate is to be taken into account at mathematical modelling of the heat and mass transfer processes in the equipment. The modelling of thermal and hydrodynamic processes in the elements of a powergenerating components is needed for developing power plants with a dissociating coolant

  6. Depleted UF6 programmatic environmental impact statement

    International Nuclear Information System (INIS)

    1997-01-01

    The US Department of Energy has developed a program for long-term management and use of depleted uranium hexafluoride, a product of the uranium enrichment process. As part of this effort, DOE is preparing a Programmatic Environmental Impact Statement (PEIS) for the depleted UF 6 management program. This report duplicates the information available at the web site (http://www.ead.anl.gov/web/newduf6) set up as a repository for the PEIS. Options for the web site include: reviewing recent additions or changes to the web site; learning more about depleted UF 6 and the PEIS; browsing the PEIS and related documents, or submitting official comments on the PEIS; downloading all or part of the PEIS documents; and adding or deleting one's name from the depleted UF 6 mailing list

  7. A definition of depletion of fish stocks

    Science.gov (United States)

    Van Oosten, John

    1949-01-01

    Attention was focused on the need of a common and better understanding of the term depletion as applied to the fisheries in order to eliminate if possible the existing inexactness of thought on the subject. Depletion has been confused at various times with at least ten different ideas associated with it but which, as has has heen pointed out, are not synonymous at all. In defining depletion we must recognize that the term represents a condition and must not he confounded with the cause (overfishing) that leads to this condition or with the symptoms that identify it. Depletion was defined as a reduction, through overfishing, in the level of abundance of the exploitable segment of a stock that prevents the realization of the maximum productive capacity.

  8. Real depletion in nodal diffusion codes

    International Nuclear Information System (INIS)

    Petkov, P.T.

    2002-01-01

    The fuel depletion is described by more than one hundred fuel isotopes in the advanced lattice codes like HELIOS, but only a few fuel isotopes are accounted for even in the advanced steady-state diffusion codes. The general assumption that the number densities of the majority of the fuel isotopes depend only on the fuel burnup is seriously in error if high burnup is considered. The real depletion conditions in the reactor core differ from the asymptotic ones at the stage of lattice depletion calculations. This study reveals which fuel isotopes should be explicitly accounted for in the diffusion codes in order to predict adequately the real depletion effects in the core. A somewhat strange conclusion is that if the real number densities of the main fissionable isotopes are not explicitly accounted for in the diffusion code, then Sm-149 should not be accounted for either, because the net error in k-inf is smaller (Authors)

  9. Fully Depleted Charge-Coupled Devices

    International Nuclear Information System (INIS)

    Holland, Stephen E.

    2006-01-01

    We have developed fully depleted, back-illuminated CCDs that build upon earlier research and development efforts directed towards technology development of silicon-strip detectors used in high-energy-physics experiments. The CCDs are fabricated on the same type of high-resistivity, float-zone-refined silicon that is used for strip detectors. The use of high-resistivity substrates allows for thick depletion regions, on the order of 200-300 um, with corresponding high detection efficiency for near-infrared and soft x-ray photons. We compare the fully depleted CCD to the p-i-n diode upon which it is based, and describe the use of fully depleted CCDs in astronomical and x-ray imaging applications

  10. Ecological considerations of natural and depleted uranium

    International Nuclear Information System (INIS)

    Hanson, W.C.

    1980-01-01

    Depleted 238 U is a major by-product of the nuclear fuel cycle for which increasing use is being made in counterweights, radiation shielding, and ordnance applications. This paper (1) summarizes the pertinent literature on natural and depleted uranium in the environment, (2) integrates results of a series of ecological studies conducted at Los Alamos Scientific Laboratory (LASL) in New Mexico where 70,000 kg of depleted and natural uranium has been expended to the environment over the past 34 years, and (3) synthesizes the information into an assessment of the ecological consequences of natural and depleted uranium released to the environment by various means. Results of studies of soil, plant, and animal communities exposed to this radiation and chemical environment over a third of a century provide a means of evaluating the behavior and effects of uranium in many contexts

  11. Plasmonic Nanoprobes for Stimulated Emission Depletion Nanoscopy.

    Science.gov (United States)

    Cortés, Emiliano; Huidobro, Paloma A; Sinclair, Hugo G; Guldbrand, Stina; Peveler, William J; Davies, Timothy; Parrinello, Simona; Görlitz, Frederik; Dunsby, Chris; Neil, Mark A A; Sivan, Yonatan; Parkin, Ivan P; French, Paul M W; Maier, Stefan A

    2016-11-22

    Plasmonic nanoparticles influence the absorption and emission processes of nearby emitters due to local enhancements of the illuminating radiation and the photonic density of states. Here, we use the plasmon resonance of metal nanoparticles in order to enhance the stimulated depletion of excited molecules for super-resolved nanoscopy. We demonstrate stimulated emission depletion (STED) nanoscopy with gold nanorods with a long axis of only 26 nm and a width of 8 nm. These particles provide an enhancement of up to 50% of the resolution compared to fluorescent-only probes without plasmonic components irradiated with the same depletion power. The nanoparticle-assisted STED probes reported here represent a ∼2 × 10 3 reduction in probe volume compared to previously used nanoparticles. Finally, we demonstrate their application toward plasmon-assisted STED cellular imaging at low-depletion powers, and we also discuss their current limitations.

  12. Skeletal Muscle Depletion and Markers for Cancer Cachexia Are Strong Prognostic Factors in Epithelial Ovarian Cancer.

    Directory of Open Access Journals (Sweden)

    Stefanie Aust

    Full Text Available Tumor cachexia is an important prognostic parameter in epithelial ovarian cancer (EOC. Tumor cachexia is characterized by metabolic and inflammatory disturbances. These conditions might be reflected by body composition measurements (BCMs ascertained by pre-operative computed tomography (CT. Thus, we aimed to identify the prognostically most relevant BCMs assessed by pre-operative CT in EOC patients.We evaluated muscle BCMs and well established markers of nutritional and inflammatory status, as well as clinical-pathological parameters in 140 consecutive patients with EOC. Furthermore, a multiplexed inflammatory marker panel of 25 cytokines was used to determine the relationship of BCMs with inflammatory markers and patient's outcome. All relevant parameters were evaluated in uni- and multivariate survival analysis.Muscle attenuation (MA-a well established BCM parameter-is an independent prognostic factor for survival in multivariate analysis (HR 2.25; p = 0.028. Low MA-reflecting a state of cachexia-is also associated with residual tumor after cytoreductive surgery (p = 0.046 and with an unfavorable performance status (p = 0.015. Moreover, MA is associated with Eotaxin and IL-10 out of the 25 cytokine multiplex marker panel in multivariate linear regression analysis (p = 0.021 and p = 0.047, respectively.MA-ascertained by routine pre-operative CT-is an independent prognostic parameter in EOC patients. Low MA is associated with the inflammatory, as well as the nutritional component of cachexia. Therefore, the clinical value of pre-operative CT could be enhanced by the assessment of MA.

  13. Depleted uranium: A DOE management guide

    International Nuclear Information System (INIS)

    1995-10-01

    The U.S. Department of Energy (DOE) has a management challenge and financial liability in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF 6 ) that are stored at the gaseous diffusion plants. The annual storage and maintenance cost is approximately $10 million. This report summarizes several studies undertaken by the DOE Office of Technology Development (OTD) to evaluate options for long-term depleted uranium management. Based on studies conducted to date, the most likely use of the depleted uranium is for shielding of spent nuclear fuel (SNF) or vitrified high-level waste (HLW) containers. The alternative to finding a use for the depleted uranium is disposal as a radioactive waste. Estimated disposal costs, utilizing existing technologies, range between $3.8 and $11.3 billion, depending on factors such as applicability of the Resource Conservation and Recovery Act (RCRA) and the location of the disposal site. The cost of recycling the depleted uranium in a concrete based shielding in SNF/HLW containers, although substantial, is comparable to or less than the cost of disposal. Consequently, the case can be made that if DOE invests in developing depleted uranium shielded containers instead of disposal, a long-term solution to the UF 6 problem is attained at comparable or lower cost than disposal as a waste. Two concepts for depleted uranium storage casks were considered in these studies. The first is based on standard fabrication concepts previously developed for depleted uranium metal. The second converts the UF 6 to an oxide aggregate that is used in concrete to make dry storage casks

  14. Seasonal total methane depletion in limestone caves

    OpenAIRE

    Waring Chris L; Hankin Stuart I; Griffith David W T; Kertesz Michael A; Kobylski Victoria; Wilson Neil L; Coleman Nicholas V; Kettlewell Graham; Zlot Robert; Bosse Michael; Bell Graham

    2017-01-01

    Methane concentration in caves is commonly much lower than the external atmosphere, yet the cave CH4 depletion causal mechanism is contested and dynamic links to external diurnal and seasonal temperature cycles unknown. Here, we report a continuous 3-year record of cave methane and other trace gases in Jenolan Caves, Australia which shows a seasonal cycle of extreme CH4 depletion, from ambient ~1,775?ppb to near zero during summer and to ~800?ppb in winter. Methanotrophic bacteria, some newly...

  15. Depleted Bulk Heterojunction Colloidal Quantum Dot Photovoltaics

    KAUST Repository

    Barkhouse, D. Aaron R.

    2011-05-26

    The first solution-processed depleted bulk heterojunction colloidal quantum dot solar cells are presented. The architecture allows for high absorption with full depletion, thereby breaking the photon absorption/carrier extraction compromise inherent in planar devices. A record power conversion of 5.5% under simulated AM 1.5 illumination conditions is reported. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  16. Nanomaterials for efficiently lowering the freezing point of anti-freeze coolants.

    Science.gov (United States)

    Hong, Haiping; Zheng, Yingsong; Roy, Walter

    2007-09-01

    In this paper, we report, for the first time, the effect of the lowered freezing point in a 50% water/50% anti-freeze coolant (PAC) or 50% water/50% ethylene glycol (EG) solution by the addition of carbon nanotubes and other particles. The experimental results indicated that the nano materials are much more efficient (hundreds fold) in lowering the freezing point than the regular ionic materials (e.g., NaCl). The possible explanation for this interesting phenomenon is the colligative property of fluid and relative small size of nano material. It is quite certain that the carbon nanotubes and metal oxide nano particles could be a wonderful candidate for the nano coolant application because they could not only increase the thermal conductivity, but also efficiently lower the freezing point of traditional coolants.

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

    International Nuclear Information System (INIS)

    Petrosyan, V.; Hovakimyan, T.; Vardanyan, M.; Khachatryan, A.; Minasyan, K.

    2010-01-01

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

  18. CFD analysis of coolant channel geometries for a tightly packed fuel rods assembly at subcritical pressure

    International Nuclear Information System (INIS)

    Guo, Rui; Oka, Yoshiaki

    2015-01-01

    Highlights: • Three coolant channel geometries are proposed for high breeding LWRs. • Thermal hydraulic performance of coolant channel geometries are analyzed. • Design ranges of PWRs and BWRs with proposed geometries are designated. • One type of geometry (Geometry B) is proved to be superior to the others. - Abstract: This paper analyzes the thermal hydraulic performance of channels with different cross sectional geometries, which were adopted by tightly packed fuel rods assembly for high breeding at operating pressure of PWR and BWR. The calculations were carried out to assess the geometrical effect of coolant channels on thermal-hydraulic parameters by using a CFD code STAR-CCM+. It is found that one type of channel geometry (geometry B) is superior to others because of broad design area of power, cladding temperature and pressure drop

  19. Management of large scale coolant channel replacement programme for Indian PHWRs

    International Nuclear Information System (INIS)

    Bhatnagar, V.K.; Chadda, S.K.; Arya, R.C.

    1994-01-01

    Coolant channel assemblies form most important core components of pressurised heavy water reactors. Zirconium alloy pressure tube which form part of coolant channel assemblies are subjected to environment of high neutron flux, high pressure and temperature. Under those operating environmental conditions, the pressure tubes material undergoes degradation of metallurgical and mechanical properties in addition to dimensional changes. The coolant channels are subjected to an in-service inspection (ISI) programme for monitoring the health particularly of the pressure tubes. The en-mass replacement of pressure tubes is needed after most of the pressure tubes show unacceptable conditions for an assured safe and reliable operation. An overview of various issues pertaining to this aspect is presented. (author). 4 figs

  20. Fuel assembly stress and deflection analysis for loss-of-coolant accident and seismic excitation

    International Nuclear Information System (INIS)

    DeMars, R.V.; Steinke, R.R.

    1975-01-01

    Babcock and Wilcox has evaluated the capability of the fuel assemblies to withstand the effects of a loss-of-coolant accident (LOCA) blowdown, the operational basis earthquake (OBE) and design basis earthquake (DBE), and the simultaneous occurrence of the DBE and LOCA. This method of analysis is applicable to all of B and W's nuclear steam system contracts that specify the skirt-supported pressure vessel. Loads during the saturated and subcooled phases of blowdown following a loss-of-coolant accident were calculated. The maximum loads on the fuel assemblies were found to be below allowable limits, and the maximum deflections of the fuel assemblies were found to be less than those that could prevent the insertion of control rods or the flow of coolant through the core. (U.S.)

  1. Duty cycle and coolant temperature trade study for tokamak power plants

    International Nuclear Information System (INIS)

    DeFreece, D.A.; Cramer, B.A.; Waganer, L.M.

    1978-01-01

    Results of the study indicate an optimum burn time of approximately 3600 seconds. The cost of electricity increases by only 3% when burn time is 2200 seconds or 5600 seconds. However, significant penalties are exacted for burn times less than 2000 seconds. For example, a burn time of 600 seconds results in a 25% cost increase when compared to a 3600 second burn. The optimum rejuvenation period is approximately 80 seconds with sharply rising penalties for reductions below 50 seconds, however, the time could be increased to 100 seconds or more with a small increase in cost of electricity. The coolant outlet temperature is optimum at 440 0 C with only a 2% penalty for operation at 400 or 500 0 C. The analysis is based on a 20% cold worked stainless steel first wall with lithium as the coolant. The effects of thermal stress cycling associated with burn and rejuvenation times along with wall loading and coolant temperature are included

  2. Magnetic forces on a ferromagnetic HT-9 first wall/blanket and coolant pipe

    International Nuclear Information System (INIS)

    Lechtenberg, T.A.; Dahms, C.; Attaya, H.; Univ. of Wisconsin, Madison)

    1984-01-01

    The GFUN 3D code was used to model the toroidal fields and determine the magnetic body forces on the STARFIRE design for coolant pipes exiting the first wall sector and first wall/blanket modules. The HT-9 coolant pipes were modeled on the basis of a square bar having the same length and material volume as the coolant pipes. The stress analysis was performed using these magnetic forces applied to a pipe of 4 meters length, 8.25 cm O.D., and 0.75 cm thickness by the MODSAP stress analysis code. For the first wall/blanket module, GFUN 3D does not allow full modeling of the complex thin-walled structure or numerous small tubes because of the element aspect ratio limitations. Therefore, to obtain three dimensional loads, a solid homogeneous equivalent structure was used

  3. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  4. Detection of coolant disturbances in the fuel elements of an LMFBR by temperature fluctuation analysis

    International Nuclear Information System (INIS)

    Weinkoetz, G.; Martin, H.; Krebs, L.

    1980-01-01

    Sodium temperature noise measurements were performed at the coolant fluid outlets of an electrically heated 169-rod bundle and also a 28-rod bundle, with different partially blocked coolant sections. On both test assemblies, a flow mixer was installed downstream of the bundle fluid exit plane. For all tests, measuring planes with three-wire thermocouples containing both steel-sodium and chromel-alumel junctions were located on the upstream and downstream sides of the flow mixer. Statistical parameters such as the root mean square (RMS) and the power spectral density (PSD) of temperature fluctuations were investigated. The influences of flow velocity, heat flux, thermocouple cut-off frequency, and different blockage sizes on these statistical parameters were analysed. Moreover, an essential result of interpretation of the experiments is that a characteristic geometrical bundle coefficient was found which indicates coolant channel disturbances only, independent of the operational conditions of the bundles such as heat flux and flow velocity. (orig.)

  5. New technique of mechanical pulse jet coolant delivery system for minimal quantity lubricant (MQL) operation

    Science.gov (United States)

    Fazli, Nik; Badrul, Hamdi, M.

    2011-01-01

    Minimum quantity lubrication (MQL) machining is one of the promising solutions to decrease the cutting fluid consumption requirements. This paper describes MQL machining in a range of lubricant consumption of 2.0ml/h, which is 10-100 times smaller than the normal consumption adopted in industries. MQL machining in this range is called pulse jet coolant delivery system in this paper. A specially designed system was used for concentrating small amounts of lubricant onto the cutting interface. The performance of concentrated spraying of lubricant in pulse jet coolant delivery system design was actual simulated and compared with others systems. The concentrated spraying of lubricant with a specially designed of system was quite effective in increasing tool life in the pulse jet coolant delivery system range.

  6. Evaluation of fire-resistant fluid lubricants for water reactor coolant pump motors

    International Nuclear Information System (INIS)

    Degueurce, B.

    1995-07-01

    Fires occurred in different reactor coolant pumps, caused by lubricant leakage and impregnation of lagging. The application of phosphate esters in steam turbine control systems eliminated fire hazard. The use of fire resistant fluids for the lubrication of reactor coolant pump motor would eliminate these fires. In 1985, it was decided to evaluate these fluids after 1977 EDF tests' results and 1980 EPRI Reports. Commercially available lubricating fluids were evaluated in the laboratory, using representative tests, to develop data to permit to select a fluid. The recommended fluid has the best results for long-term hydrolysis and showed good radiation tolerance at exposure tests at TRICASTIN power plant. The fluid should now be tests in the actual motor to confirm its properties for the lubrication of reactor coolant pumps. (author). 11 refs., 2 figs., 3 tabs

  7. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  8. Experimental stress analysis for reactor coolant pump case of PWR NPP

    International Nuclear Information System (INIS)

    Dai Xinsun

    1997-11-01

    The contents of experimental stress analysis in development for reactor coolant pump (RCP) case of PWR NPP are described. The RCP case is classified nuclear safety class 1. The electrometry and photoelastic tests provided a complete basis for the structural design and functional integrity of reactor coolant pressure foundry. The test was carried out in internal pressure, deadweight and earthquake loading and connected pipe system loading with a geometrical similarity model pump case. The detailed experimental results are provided to analyse and evaluate according to ASME code

  9. Two-phase coolant pump model of pressurized light water nuclear reactors

    International Nuclear Information System (INIS)

    Santos, G.A. dos; Freitas, R.L.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  10. Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

    International Nuclear Information System (INIS)

    Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

    2000-01-01

    The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

  11. Polonium release from an ATW burner system with liquid lead-bismuth coolant

    International Nuclear Information System (INIS)

    Li, N.; Yefimov, E.; Pankratov, D.

    1998-04-01

    The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods

  12. Membrane systems and their use in nuclear power plants. Treatment of primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kus, Pavel; Bartova, Sarka; Skala, Martin; Vonkova, Katerina [Research Centre Rez, Husinec-Rez (Czech Republic). Technological Circuits Innovation Dept.; Zach, Vaclav; Kopa, Roman [CEZ a.s., Temelin (Czech Republic). Nuclear Power Plant Temelin

    2016-03-15

    In nuclear power plants, drained primary coolant containing boric acid is currently treated in the system of evaporators and by ion exchangers. Replacement of the system of evaporators by membrane system (MS) will result in lower operating cost mainly due to lower operation temperature. In membrane systems the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of the concentrated boric acid solution together with other components, while permeate stream consists of purified water. Results are presented achieved by testing a pilot-plant unit of reverse osmosis in nuclear power plant (NPP) Temelin.

  13. Experimental investigation of thermoelectric power generation versus coolant pumping power in a microchannel heat sink

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse; Andreasen, Søren Juhl

    2012-01-01

    The coolant heat sinks in thermoelectric generators (TEG) play an important role in order to power generation in the energy systems. This paper explores the effective pumping power required for the TEGs cooling at five temperature difference of the hot and cold sides of the TEG. In addition......, the temperature distribution and the pressure drop in sample microchannels are considered at four sample coolant flow rates. The heat sink contains twenty plate-fin microchannels with hydraulic diameter equal to 0.93 mm. The experimental results show that there is a unique flow rate that gives maximum net...

  14. Coolant and ambient temperature control for chillerless liquid cooled data centers

    Science.gov (United States)

    Chainer, Timothy J.; David, Milnes P.; Iyengar, Madhusudan K.; Parida, Pritish R.; Simons, Robert E.

    2016-02-02

    Cooling control methods include measuring a temperature of air provided to a plurality of nodes by an air-to-liquid heat exchanger, measuring a temperature of at least one component of the plurality of nodes and finding a maximum component temperature across all such nodes, comparing the maximum component temperature to a first and second component threshold and comparing the air temperature to a first and second air threshold, and controlling a proportion of coolant flow and a coolant flow rate to the air-to-liquid heat exchanger and the plurality of nodes based on the comparisons.

  15. Production test IP-750 raw water as a reactor coolant. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Frymier, J.W.; Geier, R.G.

    1966-08-10

    Approximately ten years ago single-tube tests demonstrated the feasibility of using unfiltered river water as a reactor coolant from the standpoint of aluminum corrosion and film formation. However, some effluent activity penalty was indicated. Inasmuch as both current water plant operation and the characteristics of Columbia River water have changed, it was deemed appropriate to reinvestigate the use of partially treated water as a reactor coolant. This report summarizes the results of a half-reactor test carried out at F Reactor.

  16. The MIT in-pile loops for coolant chemistry and corrosion studies

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Harling, O.K.; Kohse, G.E.; Ballinger, R.G.

    1988-01-01

    A set of modular subsystems and components have been developed which can be combined in several ways to create in-pile loop facilities for experimental investigation of the effect of LWR reactor coolant chemistry on corrosion, both local and global, and on radionuclide transport and deposition. Care has been taken to scale the systems to achieve close similitude for parameters currently thought to control the phenomena of interest. Initial in-pile runs of the first of these facilities, a PWR coolant chemsitry loop, will take place soon. (author)

  17. Turbulent Coolant Dispersion in the Wake of a Turbine Vane Trailing Edge

    Science.gov (United States)

    2015-01-01

    Burlet, and D.L. Sondak. A survey of hot streak experiments and simulations. International Journal of Turbo and Jet Engines, 16(1):1–16, March 1999...dis- persion of film coolant in a turbine vane cascade”. Proceedings of ASME Turbo Expo 2014 (GT2014-25484). Additional excerpts are from Yapa SD...turbulent dis- persion of film coolant in a turbine vane cascade”. Proceedings of ASME Turbo Expo 2014 (GT2014-25484). Section 2.5 is a reformatted

  18. Reverse depletion effects and the determination of ligand density on some spherical bioparticles

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Chunxiang [Heilongjiang Bayi Agricultural University (China); Liu, Yanhui, E-mail: ionazati@itp.ac.cn; Fan, Yangtao; Liu, Yijun; Li, Qiancheng; Xu, Houqiang, E-mail: houqiangxu@yahoo.com [Guizhou University (China)

    2016-06-15

    In cell environments crowded with macromolecules, the depletion effects act and assist in the assembly of a wide range of cellular structures, from the cytoskeleton to the chromatin loop, which are well accepted. But a recent quantum dot experiment indicated that the dimensions of the receptor–ligand complex have strong effects on the size-dependent exclusion of proteins in cell environments. In this article, a continuum elastic model is constructed to resolve the competition between the dimension of the receptor–ligand complex and depletion effects in the endocytosis of a spherical virus-like bioparticle. Our results show that the depletion effects do not always assist endocytosis of a spherical virus-like bioparticle; while the dimension of the ligand–receptor complex is larger than the size of a small bioparticle in cell environments, the depletion effects do not work and reverse effects appear. The ligand density covered on the virus can be identified quantitatively.

  19. Strongly Correlated Topological Insulators

    Science.gov (United States)

    2016-02-03

    Strongly Correlated Topological Insulators In the past year, the grant was used for work in the field of topological phases, with emphasis on finding...surface of topological insulators. In the past 3 years, we have started a new direction, that of fractional topological insulators. These are materials...in which a topologically nontrivial quasi-flat band is fractionally filled and then subject to strong interactions. The views, opinions and/or

  20. Strong Cosmic Censorship

    Science.gov (United States)

    Isenberg, James

    2017-01-01

    The Hawking-Penrose theorems tell us that solutions of Einstein's equations are generally singular, in the sense of the incompleteness of causal geodesics (the paths of physical observers). These singularities might be marked by the blowup of curvature and therefore crushing tidal forces, or by the breakdown of physical determinism. Penrose has conjectured (in his `Strong Cosmic Censorship Conjecture`) that it is generically unbounded curvature that causes singularities, rather than causal breakdown. The verification that ``AVTD behavior'' (marked by the domination of time derivatives over space derivatives) is generically present in a family of solutions has proven to be a useful tool for studying model versions of Strong Cosmic Censorship in that family. I discuss some of the history of Strong Cosmic Censorship, and then discuss what is known about AVTD behavior and Strong Cosmic Censorship in families of solutions defined by varying degrees of isometry, and discuss recent results which we believe will extend this knowledge and provide new support for Strong Cosmic Censorship. I also comment on some of the recent work on ``Weak Null Singularities'', and how this relates to Strong Cosmic Censorship.

  1. The new MCNP6 depletion capability

    International Nuclear Information System (INIS)

    Fensin, M. L.; James, M. R.; Hendricks, J. S.; Goorley, J. T.

    2012-01-01

    The first MCNP based in-line Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology. (authors)

  2. Taurine depletion alters vascular reactivity in rats.

    Science.gov (United States)

    Abebe, Worku; Mozaffari, Mahmood S

    2003-09-01

    We recently showed that chronic taurine supplementation is associated with attenuation of contractile responses of rat aorta to norepinephrine and potassium chloride. However, the potential involvement of endogenous taurine in modulation of vascular reactivity is not known. Therefore, we examined the effect of beta-alanine-induced taurine depletion on the in vitro reactivity of rat aorta to selected vasoactive agents. The data indicate that both norepinephrine- and potassium-chloride-induced maximum contractile responses of endothelium-denuded aortae were enhanced in taurine-depleted rats compared with control animals. However, taurine depletion did not affect tissue sensitivity to either norepinephrine or potassium chloride. By contrast, sensitivity of the endothelium-denuded aortae to sodium nitroprusside was attenuated by taurine depletion. Similarly, taurine deficiency reduced the relaxant responses of endothelium-intact aortic rings elicited by submaximal concentrations of acetylcholine, and this effect was associated with decreased nitric oxide production. Taken together, the data suggest that taurine depletion augments contractility but attenuates relaxation of vascular smooth muscle in a nonspecific manner. Impairment of endothelium-dependent responses, which is at least in part associated with reduced nitric oxide generation, may contribute to the attenuation of the vasorelaxant responses. These vascular alterations could be of potential consequence in pathological conditions associated with taurine deficiency.

  3. Groundwater depletion embedded in international food trade

    Science.gov (United States)

    Dalin, Carole; Wada, Yoshihide; Kastner, Thomas; Puma, Michael J.

    2017-03-01

    Recent hydrological modelling and Earth observations have located and quantified alarming rates of groundwater depletion worldwide. This depletion is primarily due to water withdrawals for irrigation, but its connection with the main driver of irrigation, global food consumption, has not yet been explored. Here we show that approximately eleven per cent of non-renewable groundwater use for irrigation is embedded in international food trade, of which two-thirds are exported by Pakistan, the USA and India alone. Our quantification of groundwater depletion embedded in the world’s food trade is based on a combination of global, crop-specific estimates of non-renewable groundwater abstraction and international food trade data. A vast majority of the world’s population lives in countries sourcing nearly all their staple crop imports from partners who deplete groundwater to produce these crops, highlighting risks for global food and water security. Some countries, such as the USA, Mexico, Iran and China, are particularly exposed to these risks because they both produce and import food irrigated from rapidly depleting aquifers. Our results could help to improve the sustainability of global food production and groundwater resource management by identifying priority regions and agricultural products at risk as well as the end consumers of these products.

  4. The New MCNP6 Depletion Capability

    International Nuclear Information System (INIS)

    Fensin, Michael Lorne; James, Michael R.; Hendricks, John S.; Goorley, John T.

    2012-01-01

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.

  5. Sensibility analysis of fuel depletion using different nuclear fuel depletion codes

    Energy Technology Data Exchange (ETDEWEB)

    Martins, F.; Velasquez, C.E.; Castro, V.F.; Pereira, C.; Silva, C. A. Mello da, E-mail: felipmartins94@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: victorfariascastro@gmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Nowadays, the utilization of different nuclear codes to perform the depletion and criticality calculations has been used to simulated nuclear reactors problems. Therefore, the goal is to analyze the sensibility of the fuel depletion of a PWR assembly using three different nuclear fuel depletion codes. The burnup calculations are performed using the codes MCNP5/ORIGEN2.1 (MONTEBURNS), KENO-VI/ORIGEN-S (TRITONSCALE6.0) and MCNPX (MCNPX/CINDER90). Each nuclear code performs the burnup using different depletion codes. Each depletion code works with collapsed energies from a master library in 1, 3 and 63 groups, respectively. Besides, each code uses different ways to obtain neutron flux that influences the depletions calculation. The results present a comparison of the neutronic parameters and isotopes composition such as criticality and nuclides build-up, the deviation in results are going to be assigned to features of the depletion code in use, such as the different radioactive decay internal libraries and the numerical method involved in solving the coupled differential depletion equations. It is also seen that the longer the period is and the more time steps are chosen, the larger the deviation become. (author)

  6. Ozone depletion and chlorine loading potentials

    Science.gov (United States)

    Pyle, John A.; Wuebbles, Donald J.; Solomon, Susan; Zvenigorodsky, Sergei; Connell, Peter; Ko, Malcolm K. W.; Fisher, Donald A.; Stordal, Frode; Weisenstein, Debra

    1991-01-01

    The recognition of the roles of chlorine and bromine compounds in ozone depletion has led to the regulation or their source gases. Some source gases are expected to be more damaging to the ozone layer than others, so that scientific guidance regarding their relative impacts is needed for regulatory purposes. Parameters used for this purpose include the steady-state and time-dependent chlorine loading potential (CLP) and the ozone depletion potential (ODP). Chlorine loading potentials depend upon the estimated value and accuracy of atmospheric lifetimes and are subject to significant (approximately 20-50 percent) uncertainties for many gases. Ozone depletion potentials depend on the same factors, as well as the evaluation of the release of reactive chlorine and bromine from each source gas and corresponding ozone destruction within the stratosphere.

  7. Self-regulation, ego depletion, and inhibition.

    Science.gov (United States)

    Baumeister, Roy F

    2014-12-01

    Inhibition is a major form of self-regulation. As such, it depends on self-awareness and comparing oneself to standards and is also susceptible to fluctuations in willpower resources. Ego depletion is the state of reduced willpower caused by prior exertion of self-control. Ego depletion undermines inhibition both because restraints are weaker and because urges are felt more intensely than usual. Conscious inhibition of desires is a pervasive feature of everyday life and may be a requirement of life in civilized, cultural society, and in that sense it goes to the evolved core of human nature. Intentional inhibition not only restrains antisocial impulses but can also facilitate optimal performance, such as during test taking. Self-regulation and ego depletion- may also affect less intentional forms of inhibition, even chronic tendencies to inhibit. Broadly stated, inhibition is necessary for human social life and nearly all societies encourage and enforce it. Copyright © 2014 Elsevier Ltd. All rights reserved.

  8. Deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Parsons, P.D.; Hindle, E.D.; Mann, C.A.

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail.

  9. Alternative coolant to soluble oil in machining a mild steel material ...

    African Journals Online (AJOL)

    This paper presents the use of soybean as an alternative to soluble oil in machining a mild steel material. A detailed comparison of soluble oil and soybean oil as coolants was carried out. The coefficient of correlation (r) of soybean oil when computed was found to be 0.5, a value that fall in the range of moderate correlation.

  10. Estimation of aluminum and argon activation sources in the HANARO coolant

    International Nuclear Information System (INIS)

    Jun, Byung Jin; Lee, Byung Chul; Kim, Myong Seop

    2010-01-01

    The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant

  11. Spatial distribution of nanoparticles in PWR nanofluid coolant subjected to local nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Mirghaffari, Reza; Jahanfarnia, Gholamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2016-12-15

    Nanofluids have shown to be promising as an alternative for a PWR reactor coolant or as a safety system coolant to cover the core in the event of a loss of coolant accident. The nanoparticles distribution and neutronic parameters are intensively affected by the local boiling of nanofluid coolant. The main goal of this study was the physical-mathematical modeling of the nanoparticles distribution in the nucleate boiling of nanofluids within the viscous sublayer. Nanoparticles concentration, especially near the heat transfer surfaces, plays a significant role in the enhancement of thermal conductivity of nanofluids and prediction of CHF, Hide Out and Return phenomena. By solving the equation of convection-diffusion for the liquid phase near the heating surface and the bulk stream, the effect of heat flux on the distribution of nanoparticles was studied. The steady state mass conservation equations for liquids, vapors and nanoparticles were written for the flow boiling within the viscous sublayer adjacent the fuel cladding surface. The derived differential equations were discretized by the finite difference method and were solved numerically. It was found out that by increasing the surface heat flux, the concentration of nanoparticles increased.

  12. The effect of outflowing water coolant with supercritical parameters on a barrier

    Directory of Open Access Journals (Sweden)

    Alekseev Maksim

    2017-01-01

    Full Text Available The outflow of supercritical coolant with different initial parameters and its impact on the barrier have been numerically simulated. Spatial and axial distributions of pressure and steam quality are presented. The force acting on the barrier at different parameters of the outflow has been calculated.

  13. Simultaneous radiochemical separation of fractions of radioisotopes of primary nuclear reactor coolant water

    International Nuclear Information System (INIS)

    Csajka, M.

    1988-01-01

    Discs coated with silver iodide, lanthanum oxalate and calcium-barium sulfate were simultaneously rotated in nuclear reactor coolant water samples for radiochemical separation of iodine isotopes, a group of radioactive rare earth elements and a group of radioactive alkaline earth elements. The reproducibility of the method is discussed. (author) 10 refs.; 7 figs

  14. Partial Discharge Measurements in HV Rotating Machines in Dependence on Pressure of Coolant

    Directory of Open Access Journals (Sweden)

    I. Kršňák

    2002-01-01

    Full Text Available The influence of the pressure of the coolant used in high voltage rotating machines on partial discharges occurring in stator insulation is discussed in this paper. The first part deals with a theoretical analysis of the topic. The second part deals with the results obtained on a real generator in industrial conditions. Finally, theoretical assumptions and obtained results are compared.

  15. A New Application of Support Vector Machine Method: Condition Monitoring and Analysis of Reactor Coolant Pump

    Science.gov (United States)

    Meng, Qinghu; Meng, Qingfeng; Feng, Wuwei

    2012-05-01

    Fukushima nuclear power plant accident caused huge losses and pollution and it showed that the reactor coolant pump is very important in a nuclear power plant. Therefore, to keep the safety and reliability, the condition of the coolant pump needs to be online condition monitored and fault analyzed. In this paper, condition monitoring and analysis based on support vector machine (SVM) is proposed. This method is just to aim at the small sample studies such as reactor coolant pump. Both experiment data and field data are analyzed. In order to eliminate the noise and useless frequency, these data are disposed through a multi-band FIR filter. After that, a fault feature selection method based on principal component analysis is proposed. The related variable quantity is changed into unrelated variable quantity, and the dimension is descended. Then the SVM method is used to separate different fault characteristics. Firstly, this method is used as a two-kind classifier to separate each two different running conditions. Then the SVM is used as a multiple classifier to separate all of the different condition types. The SVM could separate these conditions successfully. After that, software based on SVM was designed for reactor coolant pump condition analysis. This software is installed on the reactor plant control system of Qinshan nuclear power plant in China. It could monitor the online data and find the pump mechanical fault automatically.

  16. Flat plate film cooling at the coolant supply into triangular and cylindrical craters

    Directory of Open Access Journals (Sweden)

    Khalatov Artem A.

    2017-01-01

    Full Text Available The results are given of the film cooling numerical simulation of three different schemes including single-array of the traditional round inclined holes, as well as inclined holes arranged in the cylindrical or triangular dimples (craters. The results of simulation showed that at the medium and high values of the blowing ratio (m > 1.0 the scheme with coolant supply into triangular craters improves the adiabatic film cooling efficiency by 1.5…2.7 times compared to the traditional array of inclined holes, or by 1.3…1.8 times compared to the scheme with coolant supply into cylindrical craters. The greater film cooling efficiency with the coolant supply into triangular craters is explained by decrease in the intensity of secondary vortex structures (“kidney” vortex. This is due to the partial destruction and transformation of the coolant jets structure interacting with front wall of the crater. Simultaneously, the film cooling uniformity is increased in the span-wise direction.

  17. Analysis of in-core-coolant temperature distributions within FFTF instrumented fuel assemblies

    International Nuclear Information System (INIS)

    Hoth, C.W.

    1982-01-01

    Instrumented fuel assembly tests are installed in the Fast Flux Test Facility and are monitoring in-core coolant temperatures. Comparison of measured temperature data to calculated temperatures from a thermal-hydraulic code confirms the validity of the thermal-hydraulic models

  18. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  19. Quantitative determination of a hydrogen impurity in a sodium coolant by hydride thermal dissociation

    Science.gov (United States)

    Ivanovskiy, M. N.; Pavlova, G. D.; Shmatko, B. A.; Milovidova, A. V.; Konovalov, E. YE.; Arnoldov, M. N.; Pleshivtsev, A. D.

    1988-01-01

    A molten sodium coolant containing hydrogen was heated in a vacuum at 450 C, and the gases generated pumped through a liquid nitrogen trap, and the H2 was then oxidized on a copper oxide substrate heated to 400 C. The accuracy of the method is 1.5 percent and the sensitivity is 1x10 to the -5 wt percent hydrogen.

  20. High-temperature process heat reactor with solid coolant and radiant heat exchange

    International Nuclear Information System (INIS)

    Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

    1984-01-01

    The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion

  1. Evaluation of Coolant Injection Procedure in the Severe Accident Management Strategy of APR1400

    International Nuclear Information System (INIS)

    Cho, Yongjin; Lim, Kukhee; Song, Sungchu; Lee, Sukho; Hwang, Taesuk

    2013-01-01

    A coolant injection strategy in the severe accident management guideline (SAMG) of APR1400 relates to immediate coolant injection into RCS (Reactor Coolant System) or injection following the recovery of secondary coolant inventory. This strategy could play important role in accident mitigation and radiological consequences. In this study, appropriateness of the strategy was evaluated using MELCOR1.8.6 and several sensitivity studies of the key parameters were performed. Analysis for APR1400 using MELCOR 1.8.6 was performed to evaluate the effectiveness of accident management strategies and the following conclusions were identified. Sequential operation of secondary and RCS injection may not be the best strategy and the simultaneous injection of secondary and RCS injection could be more preferable. At least, the RCS injection should start before complete drainage of water in the safety injection tank using mobile pumps. In this study, the effectiveness of timing of operator action has been examined and the amount of injection flowrate needs to be studied in the future

  2. Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

    Energy Technology Data Exchange (ETDEWEB)

    Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

    2012-01-15

    The Idaho National Laboratory prepared a preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 Degree-Sign C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

  3. Assessment of Structural Integrity after the Small Improvement of Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Woo; Park, Min Su; Choi, Sang Hoon; Kim, Gyung Sub; Park, Sung Ho [Korea Plant Service and Engineering, Seongnam (Korea, Republic of)

    2009-10-15

    Vibration is one of physical phenomena to affect capability and life cycles in turbo machinery. Vibration of most turbo machinery is caused by unbalanced mass, which leads turbo machinery to be damaged or failed before design life This research tried to optimize the effect on operation of Reactor Coolant Pump(RCP) due to the improvement of structure.

  4. Prototypic corium oxidation and hydrogen release during the Fuel-Coolant Interaction

    Czech Academy of Sciences Publication Activity Database

    Tyrpekl, J.; Piluso, P.; Bakardjieva, Snejana; Nižňanský, D.; Rehspringer, J.L.; Bezdička, Petr; Dugne, O.

    2015-01-01

    Roč. 75, JAN (2015), s. 210-218 ISSN 0306-4549 Institutional support: RVO:61388980 Keywords : Corium * Fuel-Coolant Interaction * Hydrogen release * Material effect * Nuclear reactor severe accident Subject RIV: CA - Inorganic Chemistry Impact factor: 1.174, year: 2015

  5. The application of release models to the interpretation of rare gas coolant activities

    International Nuclear Information System (INIS)

    Wise, C.

    1985-01-01

    Much research is carried out into the release of fission products from UO 2 fuel and from failed pins. A significant application of this data is to define models of release which can be used to interpret measured coolant activities of rare gas isotopes. Such interpretation is necessary to extract operationally relevant parameters, such as the number and size of failures in the core and the 131 I that might be released during depressurization faults. The latter figure forms part of the safety case for all operating CAGRs. This paper describes and justifies the models which are used in the ANAGRAM program to interpret CAGR coolant activities, highlighting any remaining uncertainties. The various methods by which the program can extract relevant information from the measurements are outlined, and examples are given of the analysis of coolant data. These analyses point to a generally well understood picture of fission gas release from low temperature failures. Areas of higher temperature release are identified where further research would be beneficial to coolant activity analysis. (author)

  6. Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

    2005-07-01

    Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

  7. Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

    International Nuclear Information System (INIS)

    Elina Syrjaelahti; Anitta Haemaelaeinen

    2005-01-01

    Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

  8. Probabilistic analysis of fuel pin behaviour during an eventual loss of coolant in PWR reactors

    International Nuclear Information System (INIS)

    1981-02-01

    Brief description of the development of the coolant loss incident in a pressurized water reactor and analysis of its significance for the behaviour of the fuel rods. Description of a probalistic method for estimating the effects of the accident on the fuel rods and results obtained [fr

  9. Corrosion products in the coolant circuits of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1984-01-01

    The characteristics of corrosion products formed in the primary and secondary circuits of pressurized light water nuclear power plants are first briefly recalled. The problem set by the pollution of coolants and metallic surfaces is then examined. Finally, the measures of precaution to take and the possible solutions to minimize the disturbing effects of this pollution by corrosion products are presented [fr

  10. Coolant distribution device on reactor core side of HTGR type reactor

    International Nuclear Information System (INIS)

    Masuba, Yoshitaka.

    1993-01-01

    The coolant distribution device comprises an annular main body and a pressure loss source member having a plurality of blow ports perforated at positions equally dividing the circumference of the main body. The coolant distribution device is disposed to the outer circumference of a plurality of support columns disposed between the reactor core support plate and a rectification plate at a reactor floor portion. Coolants uprising in the reactor core and flowing under heating to a reactor floor portion are blown out from main cooling system pipelines and auxiliary cooling system pipelines, passed through the gap of the reactor core support plate and the rectification plate and sealed by the pressure loss source member. Since the difference of a speed head caused by geometrical shape in the circumferential direction of a flow channel is reduced as a pressure loss, the pressure in the pressure loss source member is kept uniform. Accordingly, a uniform amount of coolants is distributed from the blow ports of the pressure loss source member to the side of the reactor core of the reactor floor portion, to cool the outer circumference of the reactor core uniformly. (I.N.)

  11. The deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Parsons, P.D.; Hindle, E.D.; Mann, C.A.

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail. (author)

  12. On Line Neutron Flux Mapping in Fuel Coolant Channels of a Research Reactor

    International Nuclear Information System (INIS)

    Barbot, Loic; Domergue, Christophe; Villard, Jean-Francois; Destouches, Christophe; Braoudakis, George; Wassink, David; Sinclair, Bradley; Osborn, John-C.; Wu, Huayou; Blandin, C.; Thevenin, Mathieu; Corre, Gwenole; Normand, Stephane

    2013-06-01

    This work deals with the on-line neutron flux mapping of the OPAL research reactor. A specific irradiation device has been set up to investigate fuel coolant channels using subminiature fission chambers to get thermal neutron flux profiles. Experimental results are compared to first neutronic calculations and show good agreement (C/E ∼0.97). (authors)

  13. Lead based alloys as high temperature nuclear reactor coolant - scope and challenges

    International Nuclear Information System (INIS)

    Borgohain, A.; Maheshwari, N.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2007-01-01

    This paper deals with the role of analytical chemistry from corrosion impurity control point of view and the identification of the major thrust areas for the development of lead based coolant technologies. This paper also deals with the activities on oxygen sensor development for the measurement of dissolved oxygen in lead-bismuth alloy. (author)

  14. Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis

    International Nuclear Information System (INIS)

    Mao, H.; Yang, B.W.; Han, B.

    2016-01-01

    Mixing vane grids (MVG) have great influence on coolant temperature field in the rod bundle. The MVG could enhance convective heat transfer between the fuel rod wall and the coolant, and promote inter-subchannel mixing at the same time. For the influence of the MVG on convective heat transfer enhancement, many experiments have been done and several correlations have been developed based on the experimental data. However, inter-subchannel mixing promotion caused by the MVG is not well estimated in subchannel analysis because the information of mixing vanes is totally missing in most subchannel codes. This paper analyzes the influence of mixing vanes on coolant temperature distribution using the improved MVG model in subchannel analysis. The coolant temperature distributions with the MVG are analyzed, and the results show that mixing vanes lead to a more uniform temperature distribution. The performances of split vane grids under different power conditions are evaluated. The results are compared with those of spacer grids without mixing vanes and some conclusions are obtained.

  15. Improvements to secondary coolant circuits of a liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Brachet, Alain.

    1981-01-01

    This invention concerns improvements to secondary coolant-systems for sodium cooled nuclear reactors. It further concerns a protective device for a free level mechanical pump which prevents any gas bubbles due to leaks of the working gas of the pump from entering the secondary system of the nuclear reactor [fr

  16. Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

  17. Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor

    Science.gov (United States)

    Puji Hastuti, Endiah; Widodo, Surip; Darwis Isnaini, M.; Geni Rina, S.; Syaiful, B.

    2018-02-01

    TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔΤ onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35°C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively.

  18. Depleted uranium hexafluoride: Waste or resource?

    International Nuclear Information System (INIS)

    Schwertz, N.; Zoller, J.; Rosen, R.; Patton, S.; Bradley, C.; Murray, A.

    1995-07-01

    The US Department of Energy is evaluating technologies for the storage, disposal, or re-use of depleted uranium hexafluoride (UF 6 ). This paper discusses the following options, and provides a technology assessment for each one: (1) conversion to UO 2 for use as mixed oxide duel, (2) conversion to UO 2 to make DUCRETE for a multi-purpose storage container, (3) conversion to depleted uranium metal for use as shielding, (4) conversion to uranium carbide for use as high-temperature gas-cooled reactor (HTGR) fuel. In addition, conversion to U 3 O 8 as an option for long-term storage is discussed

  19. Elasticity and clustering in concentrated depletion gels.

    Science.gov (United States)

    Ramakrishnan, S; Chen, Y-L; Schweizer, K S; Zukoski, C F

    2004-10-01

    X-ray scattering and rheology are employed to study the volume fraction dependence of the collective structure and elastic moduli of concentrated nanoparticle-polymer depletion gels. The nonequilibrium gel structure consists of locally densified nonfractal clusters and narrow random interfaces. The elastic moduli display a power law dependence on volume fraction with effective exponents that decrease with increasing depletion attraction strength. A microscopic theory that combines local structural information with a dynamic treatment of gelation is in good agreement with the observations.

  20. Strong Arcwise Connectedness

    OpenAIRE

    Espinoza, Benjamin; Gartside, Paul; Kovan-Bakan, Merve; Mamatelashvili, Ana

    2012-01-01

    A space is `n-strong arc connected' (n-sac) if for any n points in the space there is an arc in the space visiting them in order. A space is omega-strong arc connected (omega-sac) if it is n-sac for all n. We study these properties in finite graphs, regular continua, and rational continua. There are no 4-sac graphs, but there are 3-sac graphs and graphs which are 2-sac but not 3-sac. For every n there is an n-sac regular continuum, but no regular continuum is omega-sac. There is an omega-sac ...

  1. Abortion: Strong's counterexamples fail

    DEFF Research Database (Denmark)

    Di Nucci, Ezio

    2009-01-01

    This paper shows that the counterexamples proposed by Strong in 2008 in the Journal of Medical Ethics to Marquis's argument against abortion fail. Strong's basic idea is that there are cases--for example, terminally ill patients--where killing an adult human being is prima facie seriously morally......'s scenarios have some valuable future or admitted that killing them is not seriously morally wrong. Finally, if "valuable future" is interpreted as referring to objective standards, one ends up with implausible and unpalatable moral claims....

  2. Barriers to the Application of High-Temperature Coolants in Hybrid Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, J.S.; Staunton, M.R.; Starke, M.R.

    2006-09-30

    This study was performed by the Oak Ridge National Laboratory (ORNL) to identify practical approaches, technical barriers, and cost impacts to achieving high-temperature coolant operation for certain traction drive subassemblies and components of hybrid electric vehicles (HEV). HEVs are unique in their need for the cooling of certain dedicated-traction drive subassemblies/components that include the electric motor(s), generators(s), inverter, dc converter (where applicable), and dc-link capacitors. The new coolant system under study would abandon the dedicated 65 C coolant loop, such as used in the Prius, and instead rely on the 105 C engine cooling loop. This assessment is important because automotive manufacturers are interested in utilizing the existing water/glycol engine cooling loop to cool the HEV subassemblies in order to eliminate an additional coolant loop with its associated reliability, space, and cost requirements. In addition, the cooling of power electronic devices, traction motors, and generators is critical in meeting the U.S. Department of Energy (DOE) FreedomCAR and Vehicle Technology (FCVT) goals for power rating, volume, weight, efficiency, reliability, and cost. All of these have been addressed in this study. Because there is high interest by the original equipment manufacturers (OEMs) in reducing manufacturing cost to enhance their competitive standing, the approach taken in this analysis was designed to be a positive 'can-do' approach that would be most successful in demonstrating the potential or opportunity of relying entirely on a high-temperature coolant system. Nevertheless, it proved to be clearly evident that a few formidable technical and cost barriers exist and no effective approach for mitigating the barriers was evident in the near term. Based on comprehensive thermal tests of the Prius reported by ORNL in 2005 [1], the continuous ratings at base speed (1200 rpm) with different coolant temperatures were projected from

  3. Barriers to the Application of High-Temperature Coolants in Hybrid Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Staunton, Robert H [ORNL; Hsu, John S [ORNL; Starke, Michael R [ORNL

    2006-09-01

    This study was performed by the Oak Ridge National Laboratory (ORNL) to identify practical approaches, technical barriers, and cost impacts to achieving high-temperature coolant operation for certain traction drive subassemblies and components of hybrid electric vehicles (HEV). HEVs are unique in their need for the cooling of certain dedicated-traction drive subassemblies/components that include the electric motor(s), generators(s), inverter, dc converter (where applicable), and dc-link capacitors. The new coolant system under study would abandon the dedicated 65 C coolant loop, such as used in the Prius, and instead rely on the 105 C engine cooling loop. This assessment is important because automotive manufacturers are interested in utilizing the existing water/glycol engine cooling loop to cool the HEV subassemblies in order to eliminate an additional coolant loop with its associated reliability, space, and cost requirements. In addition, the cooling of power electronic devices, traction motors, and generators is critical in meeting the U.S. Department of Energy (DOE) FreedomCAR and Vehicle Technology (FCVT) goals for power rating, volume, weight, efficiency, reliability, and cost. All of these have been addressed in this study. Because there is high interest by the original equipment manufacturers (OEMs) in reducing manufacturing cost to enhance their competitive standing, the approach taken in this analysis was designed to be a positive 'can-do' approach that would be most successful in demonstrating the potential or opportunity of relying entirely on a high-temperature coolant system. Nevertheless, it proved to be clearly evident that a few formidable technical and cost barriers exist and no effective approach for mitigating the barriers was evident in the near term. Based on comprehensive thermal tests of the Prius reported by ORNL in 2005 [1], the continuous ratings at base speed (1200 rpm) with different coolant temperatures were projected from

  4. Coupling a system code with computational fluid dynamics for the simulation of complex coolant reactivity effects

    International Nuclear Information System (INIS)

    Bertolotto, D.

    2011-11-01

    The current doctoral research is focused on the development and validation of a coupled computational tool, to combine the advantages of computational fluid dynamics (CFD) in analyzing complex flow fields and of state-of-the-art system codes employed for nuclear power plant (NPP) simulations. Such a tool can considerably enhance the analysis of NPP transient behavior, e.g. in the case of pressurized water reactor (PWR) accident scenarios such as Main Steam Line Break (MSLB) and boron dilution, in which strong coolant flow asymmetries and multi-dimensional mixing effects strongly influence the reactivity of the reactor core, as described in Chap. 1. To start with, a literature review on code coupling is presented in Chap. 2, together with the corresponding ongoing projects in the international community. Special reference is made to the framework in which this research has been carried out, i.e. the Paul Scherrer Institute's (PSI) project STARS (Steady-state and Transient Analysis Research for the Swiss reactors). In particular, the codes chosen for the coupling, i.e. the CFD code ANSYS CFX V11.0 and the system code US-NRC TRACE V5.0, are part of the STARS codes system. Their main features are also described in Chap. 2. The development of the coupled tool, named CFX/TRACE from the names of the two constitutive codes, has proven to be a complex and broad-based task, and therefore constraints had to be put on the target requirements, while keeping in mind a certain modularity to allow future extensions to be made with minimal efforts. After careful consideration, the coupling was defined to be on-line, parallel and with non-overlapping domains connected by an interface, which was developed through the Parallel Virtual Machines (PVM) software, as described in Chap. 3. Moreover, two numerical coupling schemes were implemented and tested: a sequential explicit scheme and a sequential semi-implicit scheme. Finally, it was decided that the coupling would be single

  5. STUDY OF DRYING KINETICS OF BUCKWHEAT IN THE DRYER WITH TWISTED COOLANT FLOWS

    Directory of Open Access Journals (Sweden)

    S. T. Antipov

    2014-01-01

    Full Text Available Summary. Drying of buckwheat seeds is one of the most important stages of preparation of this raw material to the subsequent process of its processing. The nutritional value and quality indicators of the finished product depend on the mode of drying and are the result of structural, mechanical, biological and physico-mechanical transformations of substances. Technological modes of drying buckwheat seeds depend on the content of water and have a significant effect on the change of carbohydrates, protein denaturation, oxidation of lipids, changes of vitamins and organic acids. A new method of drying buckwheat and designed and constructed an experimental dryer with adjustable swirling flow of the coolant is proposed. For the study of the experiments and to determine the optimal mode of drying was used central composite rotatable uniforms - planning and selected full factorial experiment. The kinetics of drying and heat buckwheat in the device with twisted coolant flow was investigated. The influence of various parameters on the drying kinetics of buckwheat in the dryer with twisted coolant flow was invesigated. Presents the results of experimental studies, buckwheat drying in the dryer with twisted coolant flows. On the basis of experimental data and their statistical processing was obtained a mathematical model that adequately describes the process of drying buckwheat in the device with twisted coolant flow. The character of changes in the criteria optimization depending on the input factors was determined. The results of the mathematical model will be useful to a wide range of professionals involved in drying buckwheat, as well as for the calculation and design of modern drying - boiler systems.

  6. Investigating Liquid CO2 as a Coolant for a MTSA Heat Exchanger Design

    Science.gov (United States)

    Paul, Heather L.; Padilla, Sebastian; Powers, Aaron; Iacomini, Christie

    2009-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

  7. A strong comeback

    International Nuclear Information System (INIS)

    Marier, D.

    1992-01-01

    This article presents the results of a financial rankings survey which show a strong economic activity in the independent energy industry. The topics of the article include advisor turnover, overseas banks, and the increase in public offerings. The article identifies the top project finance investors for new projects and restructurings and rankings for lenders

  8. Airborne lidar measurements of surface ozone depletion over Arctic sea ice

    Directory of Open Access Journals (Sweden)

    J. A. Seabrook

    2013-06-01

    Full Text Available A differential absorption lidar (DIAL for measurement of atmospheric ozone concentration was operated aboard the Polar 5 research aircraft in order to study the depletion of ozone over Arctic sea ice. The lidar measurements during a flight over the sea ice north of Barrow, Alaska, on 3 April 2011 found a surface boundary layer depletion of ozone over a range of 300 km. The photochemical destruction of surface level ozone was strongest at the most northern point of the flight, and steadily decreased towards land. All the observed ozone-depleted air throughout the flight occurred within 300 m of the sea ice surface. A back-trajectory analysis of the air measured throughout the flight indicated that the ozone-depleted air originated from over the ice. Air at the surface that was not depleted in ozone had originated from over land. An investigation into the altitude history of the ozone-depleted air suggests a strong inverse correlation between measured ozone concentration and the amount of time the air directly interacted with the sea ice.

  9. Rb-Sr And Sm-Nd Ages, and Petrogenesis of Depleted Shergottite Northwest Africa 5990

    Science.gov (United States)

    Shih, C. Y.; Nyquist, L. E.; Reese, Y.; Irving, A. J.

    2011-01-01

    Northwest Africa (NWA) 5990 is a very fresh Martian meteorite recently found on Hamada du Draa, Morocco and was classified as an olivine-bearing diabasic igneous rock related to depleted shergottites [1]. The study of [1] also showed that NWA 5990 resembles QUE 94201 in chemical, textural and isotopic aspects, except QUE 94201 contains no olivine. The depleted shergottites are characterized by REE patterns that are highly depleted in LREE, older Sm-Nd ages of 327-575 Ma and highly LREE-depleted sources with Nd= +35+48 [2-7]. Age-dating these samples by Sm-Nd and Rb-Sr methods is very challenging because they have been strongly shocked and contain very low abundances of light rare earth elements (Sm and Nd), Rb and Sr. In addition, terrestrial contaminants which are commonly present in desert meteorites will compromise the equilibrium of isotopic systems. Since NWA 5990 is a very fresh meteorite, it probably has not been subject to significant desert weathering and thus is a good sample for isotopic studies. In this report, we present Rb-Sr and Sm-Nd isotopic results for NWA 5990, discuss the correlation of the determined ages with those of other depleted shergottites, especially QUE 94201, and discuss the petrogenesis of depleted shergottites.

  10. Serial reversal learning and acute tryptophan depletion

    NARCIS (Netherlands)

    van der Plasse, G.; Feenstra, M.G.P.

    2008-01-01

    Cognitive flexibility (i.e. the ability to adapt goal-directed behaviour in response to changed environmental demands) has repeatedly been shown to depend on the prefrontal cortex (PFC). Recent data from primate studies moreover show that depletion of prefrontal 5-HT impairs reversal learning of

  11. Global Warming: Lessons from Ozone Depletion

    Science.gov (United States)

    Hobson, Art

    2010-01-01

    My teaching and textbook have always covered many physics-related social issues, including stratospheric ozone depletion and global warming. The ozone saga is an inspiring good-news story that's instructive for solving the similar but bigger problem of global warming. Thus, as soon as students in my physics literacy course at the University of…

  12. Seasonal total methane depletion in limestone caves.

    Science.gov (United States)

    Waring, Chris L; Hankin, Stuart I; Griffith, David W T; Kertesz, Michael A; Kobylski, Victoria; Wilson, Neil L; Coleman, Nicholas V; Kettlewell, Graham; Zlot, Robert; Bosse, Michael; Bell, Graham

    2017-08-16

    Methane concentration in caves is commonly much lower than the external atmosphere, yet the cave CH 4 depletion causal mechanism is contested and dynamic links to external diurnal and seasonal temperature cycles unknown. Here, we report a continuous 3-year record of cave methane and other trace gases in Jenolan Caves, Australia which shows a seasonal cycle of extreme CH 4 depletion, from ambient ~1,775 ppb to near zero during summer and to ~800 ppb in winter. Methanotrophic bacteria, some newly-discovered, rapidly consume methane on cave surfaces and in external karst soils with lifetimes in the cave of a few hours. Extreme bacterial selection due to the absence of alternate carbon sources for growth in the cave environment has resulted in an extremely high proportion 2-12% of methanotrophs in the total bacteria present. Unexpected seasonal bias in our cave CH 4 depletion record is explained by a three-step process involving methanotrophy in aerobic karst soil above the cave, summer transport of soil-gas into the cave through epikarst, followed by further cave CH 4 depletion. Disentangling cause and effect of cave gas variations by tracing sources and sinks has identified seasonal speleothem growth bias, with implied palaeo-climate record bias.

  13. Ozone depleting substances management inventory system

    Directory of Open Access Journals (Sweden)

    Felix Ivan Romero Rodríguez

    2018-02-01

    Full Text Available Context: The care of the ozone layer is an activity that contributes to the planet's environmental stability. For this reason, the Montreal Protocol is created to control the emission of substances that deplete the ozone layer and reduce its production from an organizational point of view. However, it is also necessary to have control of those that are already circulating and those present in the equipment that cannot be replaced yet because of the context of the companies that keep it. Generally, the control mechanisms for classifying the type of substances, equipment and companies that own them, are carried in physical files, spreadsheets and text documents, which makes it difficult to control and manage the data stored in them. Method: The objective of this research is to computerize the process of control of substances that deplete the ozone layer. An evaluation and description of all process to manage Ozone-Depleting Substances (ODS, and its alternatives, is done. For computerization, the agile development methodology SCRUM is used, and for the technological solution tools and free open source technologies are used. Result: As a result of the research, a computer tool was developed that automates the process of control and management of substances that exhaust the ozone layer and its alternatives. Conclusions: The developed computer tool allows to control and manage the ozone-depleting substances and the equipment that use them. It also manages the substances that arise as alternatives to be used for the protection of the ozone layer.

  14. Nitrogen depletion in field red giants

    DEFF Research Database (Denmark)

    Masseron, T.; Lagarde, N.; Miglio, A.

    2017-01-01

    , the behaviour of nitrogen data along the evolution confirms the existence of non-canonical extramixing on the red giant branch (RGB) for all low-mass stars in the field. But more surprisingly, the data indicate that nitrogen has been depleted between the RGB tip and the red clump. This may suggest that some...

  15. [Acute tryptophan depletion in eating disorders].

    Science.gov (United States)

    Díaz-Marsa, M; Lozano, C; Herranz, A S; Asensio-Vegas, M J; Martín, O; Revert, L; Saiz-Ruiz, J; Carrasco, J L

    2006-01-01

    This work describes the rational bases justifying the use of acute tryptophan depletion technique in eating disorders (ED) and the methods and design used in our studies. Tryptophan depletion technique has been described and used in previous studies safely and makes it possible to evaluate the brain serotonin activity. Therefore it is used in the investigation of hypotheses on serotonergic deficiency in eating disorders. Furthermore, and given the relationship of the dysfunctions of serotonin activity with impulsive symptoms, the technique may be useful in biological differentiation of different subtypes, that is restrictive and bulimic, of ED. 57 female patients with DSM-IV eating disorders and 20 female controls were investigated with the tryptophan depletion test. A tryptophan-free amino acid solution was administered orally after a two-day low tryptophan diet to patients and controls. Free plasma tryptophan was measured at two and five hours following administration of the drink. Eating and emotional responses were measured with specific scales for five hours following the depletion. A study of the basic characteristics of the personality and impulsivity traits was also done. Relationship of the response to the test with the different clinical subtypes and with the temperamental and impulsive characteristics of the patients was studied. The test was effective in considerably reducing plasma tryptophan in five hours from baseline levels (76%) in the global sample. The test was well tolerated and no severe adverse effects were reported. Two patients withdrew from the test due to gastric intolerance. The tryptophan depletion test could be of value to study involvement of serotonin deficits in the symptomatology and pathophysiology of eating disorders.

  16. Strong Electroweak Symmetry Breaking

    CERN Document Server

    Grinstein, Benjamin

    2011-01-01

    Models of spontaneous breaking of electroweak symmetry by a strong interaction do not have fine tuning/hierarchy problem. They are conceptually elegant and use the only mechanism of spontaneous breaking of a gauge symmetry that is known to occur in nature. The simplest model, minimal technicolor with extended technicolor interactions, is appealing because one can calculate by scaling up from QCD. But it is ruled out on many counts: inappropriately low quark and lepton masses (or excessive FCNC), bad electroweak data fits, light scalar and vector states, etc. However, nature may not choose the minimal model and then we are stuck: except possibly through lattice simulations, we are unable to compute and test the models. In the LHC era it therefore makes sense to abandon specific models (of strong EW breaking) and concentrate on generic features that may indicate discovery. The Technicolor Straw Man is not a model but a parametrized search strategy inspired by a remarkable generic feature of walking technicolor,...

  17. Observations and Simulations of Formation of Broad Plasma Depletions Through Merging Process

    Science.gov (United States)

    Huang, Chao-Song; Retterer, J. M.; Beaujardiere, O. De La; Roddy, P. A.; Hunton, D.E.; Ballenthin, J. O.; Pfaff, Robert F.

    2012-01-01

    Broad plasma depletions in the equatorial ionosphere near dawn are region in which the plasma density is reduced by 1-3 orders of magnitude over thousands of kilometers in longitude. This phenomenon is observed repeatedly by the Communication/Navigation Outage Forecasting System (C/NOFS) satellite during deep solar minimum. The plasma flow inside the depletion region can be strongly upward. The possible causal mechanism for the formation of broad plasma depletions is that the broad depletions result from merging of multiple equatorial plasma bubbles. The purpose of this study is to demonstrate the feasibility of the merging mechanism with new observations and simulations. We present C/NOFS observations for two cases. A series of plasma bubbles is first detected by C/NOFS over a longitudinal range of 3300-3800 km around midnight. Each of the individual bubbles has a typical width of approx 100 km in longitude, and the upward ion drift velocity inside the bubbles is 200-400 m/s. The plasma bubbles rotate with the Earth to the dawn sector and become broad plasma depletions. The observations clearly show the evolution from multiple plasma bubbles to broad depletions. Large upward plasma flow occurs inside the depletion region over 3800 km in longitude and exists for approx 5 h. We also present the numerical simulations of bubble merging with the physics-based low-latitude ionospheric model. It is found that two separate plasma bubbles join together and form a single, wider bubble. The simulations show that the merging process of plasma bubbles can indeed occur in incompressible ionospheric plasma. The simulation results support the merging mechanism for the formation of broad plasma depletions.

  18. Plasmons in strong superconductors

    International Nuclear Information System (INIS)

    Baldo, M.; Ducoin, C.

    2011-01-01

    We present a study of the possible plasmon excitations that can occur in systems where strong superconductivity is present. In these systems the plasmon energy is comparable to or smaller than the pairing gap. As a prototype of these systems we consider the proton component of Neutron Star matter just below the crust when electron screening is not taken into account. For the realistic case we consider in detail the different aspects of the elementary excitations when the proton, electron components are considered within the Random-Phase Approximation generalized to the superfluid case, while the influence of the neutron component is considered only at qualitative level. Electron screening plays a major role in modifying the proton spectrum and spectral function. At the same time the electron plasmon is strongly modified and damped by the indirect coupling with the superfluid proton component, even at moderately low values of the gap. The excitation spectrum shows the interplay of the different components and their relevance for each excitation modes. The results are relevant for neutrino physics and thermodynamical processes in neutron stars. If electron screening is neglected, the spectral properties of the proton component show some resemblance with the physical situation in high-T c superconductors, and we briefly discuss similarities and differences in this connection. In a general prospect, the results of the study emphasize the role of Coulomb interaction in strong superconductors.

  19. Stratospheric ozone depletion from future nitrous oxide increases

    Directory of Open Access Journals (Sweden)

    W. Wang

    2014-12-01

    Full Text Available We have investigated the impact of the assumed nitrous oxide (N2O increases on stratospheric chemistry and dynamics using a series of idealized simulations with a coupled chemistry-climate model (CCM. In a future cooler stratosphere the net yield of NOy from N2O is shown to decrease in a reference run following the IPCC A1B scenario, but NOy can still be significantly increased by extra increases of N2O over 2001–2050. Over the last decade of simulations, 50% increases in N2O result in a maximal 6% reduction in ozone mixing ratios in the middle stratosphere at around 10 hPa and an average 2% decrease in the total ozone column (TCO compared with the control run. This enhanced destruction could cause an ozone decline in the first half of this century in the middle stratosphere around 10 hPa, while global TCO still shows an increase at the same time. The results from a multiple linear regression analysis and sensitivity simulations with different forcings show that the chemical effect of N2O increases dominates the N2O-induced ozone depletion in the stratosphere, while the dynamical and radiative effects of N2O increases are overall insignificant. The analysis of the results reveals that the ozone depleting potential of N2O varies with the time period and is influenced by the environmental conditions. For example, carbon dioxide (CO2 increases can strongly offset the ozone depletion effect of N2O.

  20. Advanced Thermal Storage for Central Receivers with Supercritical Coolants

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Bruce D.

    2010-06-15

    The principal objective of the study is to determine if supercritical heat transport fluids in a central receiver power plant, in combination with ceramic thermocline storage systems, offer a reduction in levelized energy cost over a baseline nitrate salt concept. The baseline concept uses a nitrate salt receiver, two-tank (hot and cold) nitrate salt thermal storage, and a subcritical Rankine cycle. A total of 6 plant designs were analyzed, as follows: Plant Designation Receiver Fluid Thermal Storage Rankine Cycle Subcritical nitrate salt Nitrate salt Two tank nitrate salt Subcritical Supercritical nitrate salt Nitrate salt Two tank nitrate salt Supercritical Low temperature H2O Supercritical H2O Two tank nitrate salt Supercritical High temperature H2O Supercritical H2O Packed bed thermocline Supercritical Low temperature CO2 Supercritical CO2 Two tank nitrate salt Supercritical High temperature CO2 Supercritical CO2 Packed bed thermocline Supercritical Several conclusions have been drawn from the results of the study, as follows: 1) The use of supercritical H2O as the heat transport fluid in a packed bed thermocline is likely not a practical approach. The specific heat of the fluid is a strong function of the temperatures at values near 400 °C, and the temperature profile in the bed during a charging cycle is markedly different than the profile during a discharging cycle. 2) The use of supercritical CO2 as the heat transport fluid in a packed bed thermocline is judged to be technically feasible. Nonetheless, the high operating pressures for the supercritical fluid require the use of pressure vessels to contain the storage inventory. The unit cost of the two-tank nitrate salt system is approximately $24/kWht, while the unit cost of the high pressure thermocline system is nominally 10 times as high. 3) For the supercritical fluids, the outer crown temperatures of the receiver tubes are in the range of 700 to 800 °C. At temperatures of 700 °C and above

  1. Fuel-Coolant Interaction Experiments in the TROI Facility

    Energy Technology Data Exchange (ETDEWEB)

    Min, B. T.; Hong, S. W.; Hong, S. H.; Park, I. K.; Kim, H. Y.; Song, J. H.; Kim, H. D

    2006-03-15

    A steam explosion has long been a concern in case of severe accidents in a nuclear reactor, since it might threaten the integrity of the containment. Although many studies have been performed on a steam explosion, there are still some remaining unsolved issues such as the explosivity of the real core material (corium) and the estimation of the energy conversion ratio. At the Korea Atomic Energy Research Institute (KAERI), the TROI steam explosion experiments were performed, in order to investigate the explosivity of corium. The TROI experiments were carried out to provide the experimental data for a proper estimation of a structural loading resulting from a steam explosion. These experiments were performed with prototypic materials such as ZrO{sub 2} melt and a mixture of ZrO{sub 2} and UO{sub 2} melt (corium). Total 46 tests were conducted in the TROI test series from year 2000 to the end of year 2004. The main test parameters were the variations on the composition of the melt, geometry of the interaction vessel, sub-cooling, ambient pressure, and amount of melt. Additionally the effects of an external trigger and argon environment were investigated. The main findings are that the composition, geometry, and inert gas had dominant effects on energetic steam explosions. In addition, the strength of the steam explosion was not that much strong compared to that of alumina, such as KROTOS-44. Even though efforts were made to maximize the strength of a steam explosion by increasing the amount of melt mass in water (increasing water depth), and fuel fraction (using a narrow test section), it did not work. The test results suggest that the melt of pure zirconia or eutectic corium in a wide test section leads to energetic spontaneous or triggered steam explosions, while the melt of other compositions does not.

  2. Health and environmental impact of depleted uranium

    International Nuclear Information System (INIS)

    Furitsu, Katsumi

    2010-01-01

    Depleted Uranium (DU) is 'nuclear waste' produced from the enrichment process and is mostly made up of 238 U and is depleted in the fissionable isotope 235 U compared to natural uranium (NU). Depleted uranium has about 60% of the radioactivity of natural uranium. Depleted uranium and natural uranium are identical in terms of the chemical toxicity. Uranium's high density gives depleted uranium shells increased range and penetrative power. This density, combined with uranium's pyrophoric nature, results in a high-energy kinetic weapon that can punch and burn through armour plating. Striking a hard target, depleted uranium munitions create extremely high temperatures. The uranium immediately burns and vaporizes into an aerosol, which is easily diffused in the environment. People can inhale the micro-particles of uranium oxide in an aerosol and absorb them mainly from lung. Depleted uranium has both aspects of radiological toxicity and chemical toxicity. The possible synergistic effect of both kinds of toxicities is also pointed out. Animal and cellular studies have been reported the carcinogenic, neurotoxic, immuno-toxic and some other effects of depleted uranium including the damage on reproductive system and foetus. In addition, the health effects of micro/ nano-particles, similar in size of depleted uranium aerosols produced by uranium weapons, have been reported. Aerosolized DU dust can easily spread over the battlefield spreading over civilian areas, sometimes even crossing international borders. Therefore, not only the military personnel but also the civilians can be exposed. The contamination continues after the cessation of hostilities. Taking these aspects into account, DU weapon is illegal under international humanitarian laws and is considered as one of the inhumane weapons of 'indiscriminate destruction'. The international society is now discussing the prohibition of DU weapons based on 'precautionary principle'. The 1991 Gulf War is reportedly the first

  3. Mantle depletion and metasomatism recorded in orthopyroxene in highly depleted peridotites

    DEFF Research Database (Denmark)

    Scott, James; Liu, Jingao; Pearson, D. Graham

    2016-01-01

    Although trace element concentrations in clinopyroxene serve as a useful tool for assessing the depletion and enrichment history of mantle peridotites, this is not applicable for peridotites in which the clinopyroxene component has been consumed (~ 25% partial melting). Orthopyroxene persists...

  4. International regime formation: Ozone depletion and global climate change

    International Nuclear Information System (INIS)

    Busmann, N.E.

    1994-03-01

    Two theoretical perspectives, neorealism and neoliberal institutionalism, dominate in international relations. An assessment is made of whether these perspectives provide compelling explanations of why a regime with specific targets and timetables was formed for ozone depletion, while a regime with such specificity was not formed for global climate change. In so doing, the assumptions underlying neorealism and neoliberal institutionalism are examined. A preliminary assessment is offered of the policymaking and institutional bargaining process. Patterns of interstate behavior are evolving toward broader forms of cooperation, at least with regard to global environmental issues, although this process is both slow and cautious. State coalitions on specific issues are not yet powerful enough to create a strong community of states in which states are willing to devolve power to international institutions. It is shown that regime analysis is a useful analytic framework, but it should not be mistaken for theory. Regime analysis provides an organizational framework offering a set of questions regarding the principles and norms that govern cooperation and conflict in an issue area, and whether forces independent of states exist which affect the scope of state behavior. An examination of both neorealism and neoliberal institutionalism, embodied by four approaches to regime formation, demonstrates that neither has sufficient scope to account for contextual dynamics in either the ozone depletion or global climate change regime formation processes. 261 refs

  5. Strong-coupling approximations

    International Nuclear Information System (INIS)

    Abbott, R.B.

    1984-03-01

    Standard path-integral techniques such as instanton calculations give good answers for weak-coupling problems, but become unreliable for strong-coupling. Here we consider a method of replacing the original potential by a suitably chosen harmonic oscillator potential. Physically this is motivated by the fact that potential barriers below the level of the ground-state energy of a quantum-mechanical system have little effect. Numerically, results are good, both for quantum-mechanical problems and for massive phi 4 field theory in 1 + 1 dimensions. 9 references, 6 figures

  6. Strong interaction and QFD

    International Nuclear Information System (INIS)

    Ebata, T.

    1981-01-01

    With an assumed weak multiplet structure for bosonic hadrons, which is consistent with the ΔI = 1/2 rule, it is shown that the strong interaction effective hamiltonian is compatible with the weak SU(2) x U(1) gauge transformation. Especially the rho-meson transforms as a triplet under SU(2)sub(w), and this is the origin of the rho-photon analogy. It is also shown that the existence of the non-vanishing Cabibbo angle is a necessary condition for the absence of the exotic hadrons. (orig.)

  7. Strong Coupling Holography

    CERN Document Server

    Dvali, Gia

    2009-01-01

    We show that whenever a 4-dimensional theory with N particle species emerges as a consistent low energy description of a 3-brane embedded in an asymptotically-flat (4+d)-dimensional space, the holographic scale of high-dimensional gravity sets the strong coupling scale of the 4D theory. This connection persists in the limit in which gravity can be consistently decoupled. We demonstrate this effect for orbifold planes, as well as for the solitonic branes and string theoretic D-branes. In all cases the emergence of a 4D strong coupling scale from bulk holography is a persistent phenomenon. The effect turns out to be insensitive even to such extreme deformations of the brane action that seemingly shield 4D theory from the bulk gravity effects. A well understood example of such deformation is given by large 4D Einstein term in the 3-brane action, which is known to suppress the strength of 5D gravity at short distances and change the 5D Newton's law into the four-dimensional one. Nevertheless, we observe that the ...

  8. MABEL-1. A code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.

    1978-06-01

    The MABEL-1 code has been written to investigate the deformation, of fuel pin cladding and its effects on fuel pin temperature transients during a loss-of-coolant accident. The code considers a single fuel pin with heated fuel concentric within the cladding. The fuel pin temperature distribution is evaluated using a one-dimensional conduction model with heat transfer to the coolant represented by an input set of heat transfer coefficients. The cladding deformation is calculated using the code CANSWEL, which assumes all strain to be elastic or creep and models the creep under a multi-axial stress system by a spring/dashpot combination undergoing alternate relaxation and elastic strain. (author)

  9. Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Raza, S.H.

    1993-12-01

    Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs

  10. Determination of primary flow by correlation of temperatures of the coolant

    International Nuclear Information System (INIS)

    Villanueva, Jose

    2003-01-01

    Correlation techniques are often used to assess primary coolant flow in nuclear reactors. Observable fluctuations of some physical or chemical coolant properties are suitable for this purpose. This work describes a development carried out at the National Atomic Energy Commission of Argentina (CNEA) to apply this technique to correlate temperature fluctuations. A laboratory test was performed. Two thermocouples were installed on a hydraulic loop. A stationary flow of water circulated by the mentioned loop, where a mechanical turbine type flowmeter was installed. Transit times given by the correlation flowmeter, for different flow values measured with the mechanical flowmeter, were registered and a calibration between them was done. A very good linear behavior was obtained in all the measured range. It was necessary to increase the fluctuation level by adding water at different temperatures at the measuring system input. (author)

  11. Integrated fuel-coolant interaction code: Assessment of stand-alone version 6.0

    International Nuclear Information System (INIS)

    Davis, F.J. Jr.

    1994-01-01

    This paper discusses the operational assessment of IFCI 6.0 against a small suite of experiments representative of the existing fuel-coolant interaction (FCI) database. The simulations should shakedown any obvious problems and demonstrate the functionality of the models contained within FCI for all phases of FCI phenomena. It was anticipated that these simulations should reasonably represent the experimental data. The IFCI 6.0 simulations were not expected, or required, to exactly reproduce the experimental results. IFCI 6.0 was assessed against a generic FITS-type pouring mode experiment, a FARO quenching experiment, and the IET-8 A ampersand B experiments to: (1) demonstrate that the code can reliably reproduce the results of the previous versions of the ifci code; (2) demonstrate the capability of qualitatively simulating all phases of fuel-coolant interactions (FCIs), including explosive cases, and; (3) identify shortcomings and areas for code enhancement

  12. Molten fuel/coolant interaction studies: some results obtained with the Windscale small shock tube rig

    International Nuclear Information System (INIS)

    Higham, E.J.; Vaughan, G.J.

    1978-02-01

    Experiments are described in which water has been brought into contact with various molten metals in a shock tube, thus simulating the fall of coolant into molten uranium dioxide in a postulated reactor accident. Impact velocities of the water on to the molten material were in the range 5 to 7 m/s. Shock-pulse pressures in the water column after impact and particle size distributions of the dispersed resolidified material that was recovered were measured. The proportion of dispersed material and the size of the shock pulse (by comparison with that expected from water hammer alone) have been used as criteria for the occurrence of a molten fuel/coolant interaction and such interactions of varying degrees of violence have been found for water/aluminium, water/bismuth, water/tin, over a range of temperatures from 350 0 C to 950 0 C, for water/boric oxide, but not for water/magnesium. (author)

  13. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  14. Analysis on accident transient of complete loss of forced reactor coolant flow in AP1000

    International Nuclear Information System (INIS)

    Jing Jianping; Zhang Chunming; Sun Wei; An Jieru; Jia Bin

    2014-01-01

    AP1OOO reactor is a typical third generation nuclear power plant at present in the world. The primary system of AP10OO nuclear power plant was modeled by using RELAP5/MOD 3.3 code, and the transient thermal hydraulic characteristics were analyzed under the accident sequence of complete loss of forced reactor coolant flow. The calculation results by RELAP5 code were compared with those of the COAST and LOFTRAN codes. The distributions by RELAP5 code were consistent with those of the COAST and LOFTRAN codes. The study results show that the RELAP5 model can accurately simulate the transient thermal hydraulic characteristics of AP1OOO under the accident of complete loss of forced reactor coolant flow. (authors)

  15. Analysis of radiation exposure during creep adjustment to the coolant channels at Madras Atomic Power Station

    International Nuclear Information System (INIS)

    Varadhan, R.S.; Venkataramana, K.; Kannan, R.K.; Sreekumaran Nair, B.; Chudalayandi, K.

    1994-01-01

    In pressurised heavy water reactors the coolant channels made of zircaloy-2 undergo creep deformation used intense neutron irradiation in the reactor core. In order to measure and provide for the changes in the dimensions, base line data of internal diameters, sag and length of the 306 coolant channels are measured as pre service inspection (PSI) before the reactor is loaded with fuel prior to criticality. Subsequently as part of in service inspection (ISI), axial creep of every channel is measured in every annual shutdown of the reactor and creep adjustment is done on those channels where creep expansion margin for the next one year operation is low. A study was carried out to assess the radiological impact of the job at Madras Atomic Power Station (MAPS). Various measures adopted for reducing the individual and collective doses on the job are discussed in this report. (author). 3 refs., 2 tabs

  16. Secondary seal effects in hydrostatic non-contact seals for reactor coolant pump shaft

    International Nuclear Information System (INIS)

    Fujita, T.; Koga, T.; Tanoue, H.; Hirabayashi, H.

    1987-01-01

    The paper presents a seal flow analysis in a hydrostatic non-contact seal for a PWR coolant pump shaft. A description is given of the non-contact seal for the reactor coolant pump. Results are presented for a distortion analysis of the seal ring, along with the seal flow characteristics and the contact pressure profiles of the secondary seals. The results of the work confirm previously reported findings that the seal ring distortion is sensitive to the o-ring location (which was placed between the ceramic seal face and the seal ring retainer). The paper concludes that the seal flow characteristics and the tracking performance depend upon the dynamic properties of the secondary seal. (U.K.)

  17. Vibration monitoring/diagnostic techniques, as applied to reactor coolant pumps

    International Nuclear Information System (INIS)

    Sculthorpe, B.R.; Johnson, K.M.

    1986-01-01

    With the increased awareness of reactor coolant pump (RCP) cracked shafts, brought about by the catastrophic shaft failure at Crystal River number3, Florida Power and Light Company, in conjunction with Bently Nevada Corporation, undertook a test program at St. Lucie Nuclear Unit number2, to confirm the integrity of all four RCP pump shafts. Reactor coolant pumps play a major roll in the operation of nuclear-powered generation facilities. The time required to disassemble and physically inspect a single RCP shaft would be lengthy, monetarily costly to the utility and its customers, and cause possible unnecessary man-rem exposure to plant personnel. When properly applied, vibration instrumentation can increase unit availability/reliability, as well as provide enhanced diagnostic capability. This paper reviews monitoring benefits and diagnostic techniques applicable to RCPs/motor drives

  18. Determination of mean molecular weights in organic reactor coolants. III. Differential cryoscopy with thermoelectric thermometer

    International Nuclear Information System (INIS)

    Becerro, E.; Carreira, M.

    1968-01-01

    The solubility problems raised by some components of the polymeric residue of irradiated polyphenolic coolants, which make it necessary to operate with very small samples, have been solved by means of a differential cryoscopic technique using a thermoelectric thermometer (thermal) as sensitive element. The method is based on the direct measurement of the difference between the freezing points of the investigated solution and of a reference solution whose concentration may be changed at will. The change of Δ V (mV) versus c(molal) is linear, the equivalent point being determined either analytically or graphically depending on the required accuracy. The method has been tested by measurements on pure polyphenyls, using diphenyl ether as solvent. It has been also applied to the main prospective coolants for the DON reactor. Working with 10 2 molal solutions the accuracy is better than ± 2 per cent. (Author) 2 refs

  19. Experimental study on utilization of air-borne jet sound in coolant leak detector

    International Nuclear Information System (INIS)

    Hayamizu, Y.; Kitahara, T.; Hayashi, T.; Nishimura, M.

    1975-10-01

    Studies have been undertaken to develop a new coolant leak detection method by the use of a microphone to pick up jet sound generated when pressurized high temperature water is discharged from a pressure boundary into the atmosphere. Leakage was simulated in three shapes, such as two machine-made circular holes and longitudinal and transverse slits in an inlet tube of a blowdown test facility. The measured power level of the jet sound was in agreement with theoretical values calculated from Lighthill's equation. In the study of utilization, this new method has been confirmed as applicable, and to be calculated theoretically for design on 'signal to noise ratio' evaluation. Detection of a small coolant leakage of 1 kg/sec is possible in a recirculation pump room which has large background noise from the pump if a suitable isolation wall, such as hot boxes, is installed between the monitored pipes and the pump. (auth.)

  20. Effect of coolant mixing on the static reactivity of a diluted boron slug

    International Nuclear Information System (INIS)

    Kuusisto, J.

    1999-01-01

    The effect of coolant mixing on the static reactivity of a diluted slug can be calculated with the method described in this paper. The three-dimensional boron concentration distribution based on the CFD calculations is used as an input of a static three dimensional core simulator code, which calculates the slug reactivity effect. In the examined first pump startup case with - 1200 ppm disturbance the reactor remains subcritical due to the coolant mixing even in the case of scram without the most reactive control rod. In the case of the 4 + 4 m 3 slug mixing reduces the reactivity effect by a factor of 0.32 as compared to the uniform -1200 ppm disturbance. In the case of the 16 m 3 slug filling the entire primary loop the effect of mixing reduces. The corresponding mixing factor is 0.72, but the mixing is high enough to hold the reactor subcritical

  1. LOFT transient thermal analysis for 10 inch primary coolant blowdown piping weld

    International Nuclear Information System (INIS)

    Howell, S.K.

    1978-01-01

    A flaw in a weld in the 10 inch primary coolant blowdown piping was discovered by LOFT personnel. As a result of this, a thermal analysis and fracture mechanics analysis was requested by LOFT personnel. The weld and pipe section were analyzed for a complete thermal cycle, heatup and Loss of Coolant Experiment (LOCE), using COUPLE/MOD2, a two-dimensional finite element heat conduction code. The finite element representation used in this analysis was generated by the Applied Mechanics Branch. The record of nodal temperatures for the entire transient was written on tape VSN=T9N054, and has been forwarded to the Applied Mechanics Branch for use in their mechanical analysis. Specific details and assumptions used in this analysis are found in appropriate sections of this report

  2. Analysis of coolant flow in central tube of WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Zsiros, G.; Toth, S.; Attila Aszodi, A.

    2011-01-01

    Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

  3. Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

    International Nuclear Information System (INIS)

    Hoehne, T.; Rhode, U.

    2000-01-01

    The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

  4. A Model for Molten Fuel-Coolant Interaction during Melt Slumping in a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sohal, Manohar Singh; Siefken, Larry James

    1999-10-01

    This paper describes a simple fuel melt slumping model to replace the current parametric model in SCDAP/RELAP5. Specifically, a fuel-coolant interaction (FCI) model is developed to analyze the slumping molten fuel, molten fuel breakup, heat transfer to coolant, relocation of the molten droplets, size of a partially solidified particles that settle to the bottom of the lower plenum, and melt-plenum interaction, if any. Considering our objectives, the molten fuel jet breakup model, and fuel droplets Lagrangian model as included in a code TEXAS-V with Eulerian thermal hydraulics for water and steam from SCDAP/RELAP5 were used. The model was assessed with experimental data from MAGICO-2000 tests performed at University of California at Santa Barbara, and FARO Test L-08 performed at Joint Research Center, Ispra, Italy. The comparison was found satisfactory.

  5. An Analysis of Medium Loss of Coolant Sequence for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Yong Mann

    2007-12-15

    This report contains analysis methodologies and calculation results of medium loss of Coolant sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant, 10 accident scenarios, which was predicted to have more than 10{sup -10} /ry occurrence frequency, have been analyzed as base cases for the medium loss of Coolant sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results in this report will be utilized as input data of the severe accident analysis database system.

  6. An Analysis of Large Loss of Coolant Sequence for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Yong Mann

    2007-12-15

    This report contains analysis methodologies and calculation results of Large loss of Coolant sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant, 14 accident scenarios, which was predicted to have more than 10{sup -10} /ry occurrence frequency, have been analyzed as base cases for the Large loss of Coolant sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results in this report will be utilized as input data of the severe accident analysis database system.

  7. An Analysis of Small Loss of Coolant Sequence for the Severe Accident Analysis Database

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Yong Mann

    2007-12-15

    This report contains analysis methodologies and calculation results of small loss of Coolant sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant, 10 accident scenarios, which was predicted to have more than 10{sup -9} /ry occurrence frequency, have been analyzed as base cases for the small loss of Coolant sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results in this report will be utilized as input data of the severe accident analysis database system.

  8. Analysis Of Primary Coolant Suction Side Pressure In The Delay Chamber Of The RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto

    2000-01-01

    Delay chamber is a tank to delay flow that located in the primary cooling suction side of RSG-GAS. A void occurred when operation reactor caused by too high the delta P at inlet suction pump. The condition may be avoided by using one line mode of the cooling flow. The analysis show that void volume in the delay chamber is occurred because the coolant negative pressure lowers the saturation pressure should be avoided though decreasing the delta P until about 0.1 bar at about 45 exp 0 C. Solution suggested are to use bypass flow from the spent fuel to the delay chamber. Coolant temperature can be also decreased by decreasing the power level of the reactor as well as improving the heat exchanger and cooling tower performances

  9. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 2 – Zinc

    International Nuclear Information System (INIS)

    Pease, David; LaBrier, Daniel; Ali, Amir; Blandford, Edward D.; Howe, Kerry J.

    2016-01-01

    Highlights: • Zinc release is limited to less than 1 mg/L in TSP-buffered solution under a variety of conditions (pH, temperature, zinc source). • Zinc release in high-temperature non-TSP-buffered environment is approximately 25 mg/L. • Long-term zinc release is controlled by passivation (without TSP) and zinc solubility (with TSP). • Precipitation and solubility of zinc phosphate limit the release of zinc. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of zinc from metallic zinc-bearing surfaces under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at a nuclear power generating facility. The experiments showed that in non-buffered (acidic) environments, measurable quantities of zinc are released from zinc-bearing surfaces. Precipitation and solubility of phosphate-based corrosion products, such as zinc phosphate, limit the release of zinc from zinc-bearing surfaces. These experiments have found that under a variety of conditions, including variations of temperature, pH, and across different zinc-bearing surfaces, the release of zinc into solution is limited to <1 mg/L when phosphate is present. When phosphate is not present, zinc release is instead bounded by a markedly higher saturation limit which is a strong function of the solution temperature.

  10. Experimental investigation of material chemical effects on emergency core cooling pump suction filter performance after loss of coolant accident

    International Nuclear Information System (INIS)

    Park, Jong Woon; Park, Byung Gi; Kim, Chang Hyun

    2009-01-01

    Integral tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for 30 days after a loss of coolant accident. A test rig with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber and the amounts of reactor building materials are scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of calcium-silicate, reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON TM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

  11. Reactor fuel depletion benchmark of TINDER

    International Nuclear Information System (INIS)

    Martin, W.J.; Oliveira, C.R.E. de; Hecht, A.A.

    2014-01-01

    Highlights: • A reactor burnup benchmark of TINDER, coupling MCNP6 to CINDER2008, was performed. • TINDER is a poor candidate for fuel depletion calculations using its current libraries. • Data library modification is necessary if fuel depletion is desired from TINDER. - Abstract: Accurate burnup calculations are key to proper nuclear reactor design, fuel cycle modeling, and disposal estimations. The TINDER code, originally designed for activation analyses, has been modified to handle full burnup calculations, including the widely used predictor–corrector feature. In order to properly characterize the performance of TINDER for this application, a benchmark calculation was performed. Although the results followed the trends of past benchmarked codes for a UO 2 PWR fuel sample from the Takahama-3 reactor, there were obvious deficiencies in the final result, likely in the nuclear data library that was used. Isotopic comparisons versus experiment and past code benchmarks are given, as well as hypothesized areas of deficiency and future work

  12. Depleted uranium hexafluoride: Waste or resource?

    Energy Technology Data Exchange (ETDEWEB)

    Schwertz, N.; Zoller, J.; Rosen, R.; Patton, S. [Lawrence Livermore National Lab., CA (United States); Bradley, C. [USDOE Office of Nuclear Energy, Science, Technology, Washington, DC (United States); Murray, A. [SAIC (United States)

    1995-07-01

    the US Department of Energy is evaluating technologies for the storage, disposal, or re-use of depleted uranium hexafluoride (UF{sub 6}). This paper discusses the following options, and provides a technology assessment for each one: (1) conversion to UO{sub 2} for use as mixed oxide duel, (2) conversion to UO{sub 2} to make DUCRETE for a multi-purpose storage container, (3) conversion to depleted uranium metal for use as shielding, (4) conversion to uranium carbide for use as high-temperature gas-cooled reactor (HTGR) fuel. In addition, conversion to U{sub 3}O{sub 8} as an option for long-term storage is discussed.

  13. The depletion of the stratospheric ozone layer

    International Nuclear Information System (INIS)

    Sabogal Nelson

    2000-01-01

    The protection of the Earth's ozone layer is of the highest importance to mankind. The dangers of its destruction are by now well known. The depletion of that layer has reached record levels. The Antarctic ozone hole covered this year a record area. The ozone layer is predicted to begin recovery in the next one or two decades and should be restored to pre-1980 levels by 2050. This is the achievement of the regime established by the 1985 Vienna Convention for the Protection of the Ozone Layer and the 1987 Montreal Protocol on Substances that Deplete the Ozone Layer. The regime established by these two agreements has been revised, and made more effective in London (1990), Copenhagen (1992), Vienna (1995), and Beijing (1999)

  14. Depleted uranium plasma reduction system study

    International Nuclear Information System (INIS)

    Rekemeyer, P.; Feizollahi, F.; Quapp, W.J.; Brown, B.W.

    1994-12-01

    A system life-cycle cost study was conducted of a preliminary design concept for a plasma reduction process for converting depleted uranium to uranium metal and anhydrous HF. The plasma-based process is expected to offer significant economic and environmental advantages over present technology. Depleted Uranium is currently stored in the form of solid UF 6 , of which approximately 575,000 metric tons is stored at three locations in the U.S. The proposed system is preconceptual in nature, but includes all necessary processing equipment and facilities to perform the process. The study has identified total processing cost of approximately $3.00/kg of UF 6 processed. Based on the results of this study, the development of a laboratory-scale system (1 kg/h throughput of UF6) is warranted. Further scaling of the process to pilot scale will be determined after laboratory testing is complete

  15. Improvements in EBR-2 core depletion calculations

    International Nuclear Information System (INIS)

    Finck, P.J.; Hill, R.N.; Sakamoto, S.

    1991-01-01

    The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs

  16. Depletion zones and crystallography on pinched spheres

    Science.gov (United States)

    Chen, Jingyuan; Xing, Xiangjun; Yao, Zhenwei

    2018-03-01

    Understanding the interplay between ordered structures and substrate curvature is an interesting problem with versatile applications, including functionalization of charged supramolecular surfaces and modern microfluidic technologies. In this work, we investigate the two-dimensional packing structures of charged particles confined on a pinched sphere. By continuously pinching the sphere, we observe cleavage of elongated scars into pleats, proliferation of disclinations, and subsequently, emergence of a depletion zone at the negatively curved waist that is completely void of particles. We systematically study the geometrics and energetics of the depletion zone, and reveal its physical origin as a finite size effect, due to the interplay between Coulomb repulsion and concave geometry of the pinched sphere. These results further our understanding of crystallography on curved surfaces, and have implications in design and manipulation of charged, deformable interfaces in various applications.

  17. Molten-Salt Depleted-Uranium Reactor

    OpenAIRE

    Dong, Bao-Guo; Dong, Pei; Gu, Ji-Yuan

    2015-01-01

    The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results...

  18. The ultimate disposition of depleted uranium

    Energy Technology Data Exchange (ETDEWEB)

    Lemons, T.R. [Uranium Enrichment Organization, Oak Ridge, TN (United States)

    1991-12-31

    Depleted uranium (DU) is produced as a by-product of the uranium enrichment process. Over 340,000 MTU of DU in the form of UF{sub 6} have been accumulated at the US government gaseous diffusion plants and the stockpile continues to grow. An overview of issues and objectives associated with the inventory management and the ultimate disposition of this material is presented.

  19. Optical assessment of phytoplankton nutrient depletion

    DEFF Research Database (Denmark)

    Heath, M.R.; Richardson, Katherine; Kiørboe, Thomas

    1990-01-01

    The ratio of light absorption at 480 and 665 nm by 90% acetone extracts of marine phytoplankton pigments has been examined as a potential indicator of phytoplankton nutritional status in both laboratory and field studies. The laboratory studies demonstrated a clear relationship between nutritiona......-replete and nutrient-depleted cells. The field data suggest that the absorption ratio may be a useful indicator of nutritional status of natural phytoplankton populations, and can be used to augment the interpretation of other data....

  20. Heatstroke Pathophysiology: The Energy Depletion Model

    Science.gov (United States)

    1989-06-12

    metabolic acidosis suggested preexisting whole body K* deficiency. In Kt-depleted dogs (49), muscle weakness occurred when animals had lost...and A.C. Issekutz. Lactate metabolism in resting and exercising dogs . J. ADDI. Physiol. 40:312-319, 1976. 43. Jacobs, I. Brod lactate: Implications...31 50. Kreisberg, R.A., L.F. Pennington, and B.R. Boshell. Lactate turnover and gluconeogenesis in normal and obese humans. Diabetes 19:53-63, 1970

  1. Carbon sequestration in depleted oil shale deposits

    Science.gov (United States)

    Burnham, Alan K; Carroll, Susan A

    2014-12-02

    A method and apparatus are described for sequestering carbon dioxide underground by mineralizing the carbon dioxide with coinjected fluids and minerals remaining from the extraction shale oil. In one embodiment, the oil shale of an illite-rich oil shale is heated to pyrolyze the shale underground, and carbon dioxide is provided to the remaining depleted oil shale while at an elevated temperature. Conditions are sufficient to mineralize the carbon dioxide.

  2. MABEL-2: a code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.; Haste, T.J.

    1982-04-01

    MABEL can be used to determine the cladding deformation in a PWR during a LOCA. It takes the results of calculations from other codes to define the initial fuel condition and the transient whole core thermal-hydraulic behaviour. The use of MABEL with input data appropriate to different regions of a reactor core allows an overall picture of coolant channel blockage within the core to be obtained. (U.K.)

  3. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    International Nuclear Information System (INIS)

    Siamidis, John; Mason, Lee

    2006-01-01

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed excel analytical model, HRS O pt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids

  4. Main-coolant-pump shaft-seal guidelines. Volume 3. Specification guidelines. Final report

    International Nuclear Information System (INIS)

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and criteria to aid in the generation of procurement specifications for Main Coolant Pump Shaft Seals. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies, a review of pump and shaft seal literature and discussions with pump and seal designers. This report is preliminary in nature and could be expanded and finalized subsequent to completion of further design, test and evaluation efforts

  5. Main-coolant-pump shaft-seal guidelines. Volume 2. Operational guidelines. Final report

    International Nuclear Information System (INIS)

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and criteria for improving main coolant pump shaft seal operational reliability. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies. Usage procedures/practices and operational environment influence on seal life and reliability from the most recent such survey are summarized. The shaft seal and its auxiliary supporting systems are discussed both from technical and operational related viewpoints

  6. Recent developments in coolant systems for Indus Accelerator Complex at RRCAT, Indore

    International Nuclear Information System (INIS)

    Nanda, Dipankar; Tiwari, Bablu; Pandey, R.M.

    2015-01-01

    Scarcity of fresh water forces mankind to explore other possible water sources that can meet the increasing demand of coolants in industries, R and D sectors and other establishments where water is used as coolant. It also becomes a challenge for water chemist to control water chemistry to keep the equipments/devices intact during its operation using water as coolant. Deionised (DI) and soft water have been used as coolants for Indus Accelerator Complex, RRCAT, Indore. DI water is produced and its quality is maintained either by conventional ion exchange method or a hybrid method of membrane separation and ion exchange technique. This requires handling of corrosive chemicals, manpower, space for plant installation, and a long array of water treatment units. CSL has implemented the idea of rain water harvesting to produce DI water after systematic studies in laboratory. The concerning issues are reduced to almost one-fourth by using rain water to produce DI water. The harvesting system has been in use for last three years. Heat is dissipated into air by evaporation of soft water in cooling tower. Requirement of soft water makeup has been estimated to be about 40,000 ltrs. / day (max.) if the machine is operated at its designed specifications. Non-availability of soft water (which circulates in open loop) may lead to shut down like situation and looking for alternate source becomes quite essential. Laboratory studies (water analysis and treatment) on sewage water (available 1,00,000 ltrs/day) from RRCAT colony as a possible source of producing soft water show promising result. (author)

  7. MABEL-2: a code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.

    1982-04-01

    MABEL-2 has been developed to predict the extent of cladding deformation in PWR fuel rods during a loss of coolant accident. The user notes describe how to run MABEL. They include case preparation and input data, the job control language, a description of the output tables available, and aids to debugging. The input data and results for two sample cases are given. (U.K.)

  8. Spectrographic determination of metallic impurities in organic coolants for nuclear reactors

    International Nuclear Information System (INIS)

    Martin Munoz, M.; Alvarez Gonzalez, F.

    1969-01-01

    A spectrochemical method for determining metallic impurities in organic coolants for nuclear reactors is given. The organic matter in solid samples is eliminated by controlled distillation and dry ashing in the presence of magnesium oxide as carrier. Liquid, samples are vacuum distillated. The residue is analyzed by carrier distillation and by total burning techniques. The analytical results are discussed and compared with those obtained destroying the organic matter without carrier and using the copper spark technique. (Author) 12 refs

  9. Attenuation of Vane-Rotor Shock Interactions with Pulsating Coolant Flows

    Science.gov (United States)

    2012-03-01

    consists on, pulsating coolant blowing through the trailing edge of the airfoils . This paper presents an unprecedented experimental and numerical...A transonic turbine airfoil was experimentally and numerically tested to resolve the effects of pulsating cooling The model was tested at four Mach...interactions are one of the major sources of losses in supersonic airfoils . Additionally, vane shocks cause large pressure fluctuations that may result in

  10. Hard alloys testing-machine for values of PWR primary coolant circuits

    International Nuclear Information System (INIS)

    Campan, J.L.; Sauze, A.

    1980-01-01

    Testing of valve parts or material used in valve fabrication and particularly seizing conditions in friction of plane surfaces coated with hard alloys of the type stellite. The testing equipment called Marguerite is composed of a hot pressurized water loop in conditions similar to PWR primary coolant circuits (320 0 C, 150 bars) and a testing-machine with measuring instruments. Testing conditions and samples are described [fr

  11. Description of steam condensation phenomena during the loss-of-coolant accident

    International Nuclear Information System (INIS)

    McCauley, E.W.; Holman, G.S.; Aust, E.; Furst, H.; Schwan, H.; Vollbrandt, J.

    1981-01-01

    Study of results from the full scale multivent pressure suppression experiment conducted by the GKSS Laboratory has developed an improved understanding of the dynamic, oscillatory steam condensation events and related loading functions which occur during the hypothetical loss-of-coolant accident in a boiling water nuclear reactor. Due to the unique measurements systems which combines both cinematic and digital data, qualified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena has been obtained

  12. Effect of tungsten-187 in primary coolant on dose rate build-up in Vandellos 2

    International Nuclear Information System (INIS)

    Fernandez Lillo, E.; Llovet, R.; Boronat, M.

    1994-01-01

    The present work proposes a relationship between the Cobalt-60 piping deposited activity and the relatively high levels of Tungsten-187 in the coolant of Vandellos 2. The conclusions of this work can be applicable to other plants, since it proposes a tool to estimate and quantify the contribution of stellite to the generation of Cobalt-60 and the radiation dose build-up. (authors). 7 figs., 6 refs

  13. Preliminary assessment of water-based nano-fluids for use as coolants in PWRs

    International Nuclear Information System (INIS)

    Jacopo Buongiorno

    2005-01-01

    Full text of publication follows: The impact of using water-based fluids with small additions (<2% vol.) of nano-sized (10-100 nm) particle populations as coolants for current and advanced PWRs is evaluated. Such 'engineered' fluids (known as nano-fluids) are attractive because the presence of the nano-particles enhances energy transport considerably. As a result, nano-fluids are known to have (i) higher thermal conductivity than water (up to 20% depending on nano-particle material, size and volumetric fraction), (ii) higher heat transfer coefficients (up to 40%), (iii) higher CHF (up to 300% in pool boiling), and (iv) comparable pressure drop. Furthermore, nano-fluids appear to be very stable suspensions with little or no sedimentation, because of the small size of the dispersed particles and their typically low volumetric fractions. The ultimate objective of this work is to assess whether existing PWRs could be retro-fitted with a water-based nano-fluid coolant, to increase safety margins, reduce stored energy, and/or allow for power up-rates. Also, advanced PWRs could be designed with nano-fluids. The linear heat generation rate in PWRs is limited by a) fuel centerline melting, b) cladding overheating (CHF), and c) stored energy release following a large-break LOCA. Mechanisms b) and c) are usually the most limiting. For given geometry and linear power, it is obvious that the core with the nano-fluid coolant will have higher margins to CHF and LOCA limits. Conversely, for given margins, a higher linear power can be accommodated by the nano-fluid-cooled core. Standard thermal-hydraulic models for the PWR hot fuel pin (including a RELAP model for the LOCA) have been used to quantify the benefit of using nano-fluid coolants on the performance of a PWR. (author)

  14. TMI-2 [Three Mile Island Unit 2] primary coolant mass flowrate data report

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1986-12-01

    This is a report on the preparation of data from the TMI-2 primary coolant mass flowrate meters for inclusion into the TMI Data Base. The sources of the as-recorded data are discussed, and a description of the instrument is given. An explanation is given of how corrections were made to the as-recorded data and how the uncertainties were calculated. The identifiers attached to each data set in the TMI Data Base are given

  15. Main-coolant-pump shaft-seal reliability investigation. Interim report

    International Nuclear Information System (INIS)

    Fair, C.E.; Marsi, J.A.; Greer, A.O.

    1982-09-01

    This report contains the results of a survey of reactor coolant pump shaft seal reliability. The survey sample is representatively large (approx. = 27% of total US commercial plant population) and includes the three industry seal suppliers (Bingham-Williamette, Byron Jackson, and Westinghouse). Operationally incurred/induced problems and seal redesign parameters are identified. Failure hypotheses in the form of fault trees have been developed to describe the failure mechanisms. Recommendations are made for seal reliability improvement

  16. Main-coolant-pump shaft-seal guidelines. Volume 1. Maintenance-manual guidelines. Final report

    International Nuclear Information System (INIS)

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and a listing of information and data which should be included in maintenance manuals and procedures for Main Coolant Pump Shaft Seals. The noted guidelines and data listing are developed from EPRI sponsored nuclear plant seal operating experience studies. The maintenance oriented results of the most recent such study is summarized. The shaft seal and its auxiliary supporting systems are discussed from both technical and maintenance related viewpoints

  17. Experimental Studies on Breakup and Fragmentation Behavior of Molten Tin and Coolant Interaction

    OpenAIRE

    Li, Yankai; Wang, Zefeng; Lin, Meng; Zhong, Mingjun; Zhou, Yueshan; Yang, Yanhua

    2017-01-01

    Jet breakup and fragmentation behavior significantly affect the likelihood (and ultimate strength) of steam explosion, but it is very challenging to assess the potential damage to reactor cavity due to general lack of knowledge regarding jet breakup phenomena. In this study, the METRIC (mechanism study test apparatus for melt-coolant interaction) was launched at Shanghai Jiao Tong University to investigate FCI physics. The first five tests on molten tin and water interactions are analyzed in ...

  18. Primary Coolant pH Control for Soluble Boron-Free PWRs

    International Nuclear Information System (INIS)

    Cheon, Yang Ho; Lee, Nam Yeong; Park, Byeong Ho; Park, Seong Chan; Kim, Eun Kee

    2015-01-01

    These should be considered when evaluating and designing the operating pH program for nuclear power plants. This paper discusses the advanced water chemistry strategies to keep pace with the recent global trends related to pH control in the primary water system for soluble boron pressurized water reactor (PWR) plants. Finally, the objective of this work is to study primary coolant pH control for soluble boron-free PWR plants. This paper reviewed the advanced water chemistry strategies to keep pace with the recent global trends related to pH control in the primary water chemistry system for soluble boron PWR plants. The new chemistry trend for the primary coolant is towards adaption of the constant and elevated chemistry. Finally, this work studied primary coolant pH control for soluble boron-free PWR plants. The ammonia-based water chemistry related to pH control for boron-free PWR plants was discussed. The ammonia-based water chemistry is not recommended to avoid fluctuation of the pH value by ammonia radiolysis and to reduce C-14 production in reactor coolant from reaction with dissolved nitrogen. Also, the potassium-based water chemistry related to pH control for boron-free PWR plants was discussed. KOH has a potential as an alternative pH control agent for soluble boron-free PWR plants. The potassium-based water chemistry related to pH control is recommended for boron-free operation as follows

  19. Fine element (F.E.) modelling of hydrogen migration and blister formation in PHWR coolant channels

    International Nuclear Information System (INIS)

    Prasad, P.S.; Dutta, B.K.; Sinha, R.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1995-01-01

    The formation of a cold spot in pressure tube due to its contact with calandria tube of PHWR coolant results in the migration of Hydrogen in pressure tube towards contact zone from its surrounding material. A 3-D finite element code SPARSH is developed to model the hydrogen redistribution and consequent hydride blister formation due to thermal and Hydrogen concentration gradients. In the present paper, the details and performance of this code are presented. (author). 6 refs., 2 figs

  20. Additional requirements for leak-before-break application to primary coolant piping in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G. [AIB Vincotte Nuclear, Brussels (Belgium)

    1997-04-01

    Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbers at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.

  1. Ultrasonic in-service inspection of main coolant pump bowl welds. Part 1. Examination technique

    International Nuclear Information System (INIS)

    Dombret, P.; Delaide, M.; Maes, G.

    1992-01-01

    The paper reports on a new ultrasonic testing technique that permits volumetric examination of PWR Main Coolant Pump casing welds from the outside of the pump bowl. The following describes the past and current evolution of the in-service inspection practice, in the light of the ASME Code requirements, and outlines the main features of the ultrasonic technique, which is based on large diameter focused beam probes. A review of the tests applied to qualify the method is also given

  2. Dynamics of UF6 Desublimation with the Influence of Tank Geometry for Various Coolant Temperature

    Directory of Open Access Journals (Sweden)

    Orlov Аleksey А.

    2016-01-01

    Full Text Available Mathematical model of UF6 desublimation in a vertical immersion tank is presented in the article. Results of calculations of the filling dynamics of the tanks with 1m3 volume at various coolant temperatures, with and without ellipticity of the end walls are given. It is shown that allowance for the ellipticity of the end walls of the tanks leads to a significant increase in the time of desublimation of UF6.

  3. Dynamics of UF6 Desublimation with the Influence of Tank Geometry for Various Coolant Temperature

    OpenAIRE

    Orlov Аleksey А.; Tsimbalyuk Alexandr F.; Malyugin Roman V.; Glazunov Anatolij A.

    2016-01-01

    Mathematical model of UF[6] desublimation in a vertical immersion tank is presented in the article. Results of calculations of the filling dynamics of the tanks with 1m3 volume at various coolant temperatures, with and without ellipticity of the end walls are given. It is shown that allowance for the ellipticity of the end walls of the tanks leads to a significant increase in the time of desublimation of UF[6].

  4. Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

    Energy Technology Data Exchange (ETDEWEB)

    Soli T. Khericha

    2006-09-01

    This report presents preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T&FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420oC. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation.

  5. Depleted uranium: Metabolic disruptor?; Uranium appauvri: perturbateur metabolique?

    Energy Technology Data Exchange (ETDEWEB)

    Souidi, Maamar; Dublineau, Isabelle; Lestaevel, Philippe [Institut de Radioprotection et de Surete Nucleaire - IRSN, Direction de la radioprotection de l' homme, Laboratoire de radiotoxicologie experimentale, Service de radiobiologie et d' epidemiologie, BP 17, 92262 Fontenay-aux-Roses cedex (France)

    2011-11-15

    The presence of uranium in the environment can lead to long-term contamination of the food chain and of water intended for human consumption and thus raises many questions about the scientific and societal consequences of this exposure on population health. Although the biological effects of chronic low-level exposure are poorly understood, results of various recent studies show that contamination by depleted uranium (DU) induces subtle but significant biological effects at the molecular level in organs including the brain, liver, kidneys and testicles. For the first time, it has been demonstrated that DU induces effects on several metabolic pathways, including those metabolizing vitamin D, cholesterol, steroid hormones, acetylcholine and xenobiotics. This evidence strongly suggests that DU might well interfere with many metabolic pathways. It might thus contribute, together with other man-made substances in the environment, to increased health risks in some regions. (authors)

  6. LIGO: The strong belief

    CERN Multimedia

    Antonella Del Rosso

    2016-01-01

    Twenty years of designing, building and testing a number of innovative technologies, with the strong belief that the endeavour would lead to a historic breakthrough. The Bulletin publishes an abstract of the Courier’s interview with Barry Barish, one of the founding fathers of LIGO.   The plots show the signals of gravitational waves detected by the twin LIGO observatories at Livingston, Louisiana, and Hanford, Washington. (Image: Caltech/MIT/LIGO Lab) On 11 February, the Laser Interferometer Gravitational-Wave Observatory (LIGO) and Virgo collaborations published a historic paper in which they showed a gravitational signal emitted by the merger of two black holes. These results come after 20 years of hard work by a large collaboration of scientists operating the two LIGO observatories in the US. Barry Barish, Linde Professor of Physics, Emeritus at the California Institute of Technology and former Director of the Global Design Effort for the Internat...

  7. Mixing requirements for the limiting fuel-coolant interactions in liquid metal fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lenz, Jr, W F

    1976-11-01

    An estimation of the mixing requirements for the limiting fuel-coolant interactions in two specific liquid metal cooled fast reactors, the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR), has been undertaken. The mixing requirements were represented in terms of the limiting mixing time constants. These constants were determined with the Argonne parametric FCI Computer Model for a range of core involvements. Specifically, fuel masses used ranged from as low as one-seventh of the core to a full core involvement. In general, conservative values for additional FCI input parameters were assumed such that the results would be conservative. With the results in hand, several mechanisms were investigated to determine what limiting effects they could have on the mixing rates of the fuel and coolant during an FCI. The energy requirements for mixing were investigated. The results, however, provided no limiting effects. A solidification limited fragmentation model was also investigated. Although this model provided no absolute limiting effects, it did show that fuel particle sizes of a certain size could indeed limit the fuel-coolant mixing rates. Additionally, the limiting effects were found to be much less significant for UC fuel. The third mechanism that was investigated concerned the limiting effects of the finite fuel release rates as a result of TOP accidents in the FFTF. Equivalent mixing time constants based on the fuel release rates were shown to be greater than the limiting values. Thus, this mechanism was shown to be limiting for the particular accident sequence investigated.

  8. Fuel, structural material and coolant for an advanced fast micro-reactor

    International Nuclear Information System (INIS)

    Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

    2011-01-01

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

  9. Estimation of the mechanical behavior of irradiated coolant channels at a nuclear plant for its decomissing

    International Nuclear Information System (INIS)

    Piquin, Ruben; Zanni, Pablo

    2003-01-01

    The widespread replacement of reactor internals generates a substantial volume of active material.It is essential to work with these components at least in a partial way before the next planned stop.Due to the fact that the reactor internals pool and the storage pools for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools.Basically the processed waste will be put in baskets at the bottom of the storage pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6.Having planned the constructive and operative solutions, the mechanical characterization of the different parts of the channel tube remains to be done.In the present paper, the necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and for the zircaloy-4 section, starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation

  10. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  11. The high-temperature sodium coolant technology in nuclear power installations for hydrogen power engineering

    Science.gov (United States)

    Kozlov, F. A.; Sorokin, A. P.; Alekseev, V. V.; Konovalov, M. A.

    2014-05-01

    In the case of using high-temperature sodium-cooled nuclear power installations for obtaining hydrogen and for other innovative applications (gasification and fluidization of coal, deep petroleum refining, conversion of biomass into liquid fuel, in the chemical industry, metallurgy, food industry, etc.), the sources of hydrogen that enters from the reactor plant tertiary coolant circuit into its secondary coolant circuit have intensity two or three orders of magnitude higher than that of hydrogen sources at a nuclear power plant (NPP) equipped with a BN-600 reactor. Fundamentally new process solutions are proposed for such conditions. The main prerequisite for implementing them is that the hydrogen concentration in sodium coolant is a factor of 100-1000 higher than it is in modern NPPs taken in combination with removal of hydrogen from sodium by subjecting it to vacuum through membranes made of vanadium or niobium. Numerical investigations carried out using a diffusion model showed that, by varying such parameters as fuel rod cladding material, its thickness, and time of operation in developing the fuel rods for high-temperature nuclear power installations (HT NPIs) it is possible to exclude ingress of cesium into sodium through the sealed fuel rod cladding. However, if the fuel rod cladding loses its tightness, operation of the HT NPI with cesium in the sodium will be unavoidable. Under such conditions, measures must be taken for deeply purifying sodium from cesium in order to minimize the diffusion of cesium into the structural materials.

  12. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Gene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-04-05

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  13. Primary coolant sampling for activated corrosion product studies at Hanford N Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, D.B.

    1985-01-31

    A special system for sampling primary coolant at N Reactor during operation has been constructed and operated from 1977 to 1983. The basic criteria and design for solving the difficult problem of getting representative samples have been presented; this report details how the instrumentation was configured and sampling was done. Equipment and procedures were put together to allow one person to enter a radiation zone, check on 5 monitoring instruments, operate two batch instruments, gather five partitioned samples, record 26 pieces of information, annotate a strip chart and leave the zone in 30 minutes while expending 10 mRem of exposure. Additionally, the reduction of the samples' analysis, digitization of strip chart information and storage of all data on data management systems is maintained. As built, the system provides 0.3 to 1.0 gpm streams of coolant from upstream and downstream of a steam generator. The streams are cooled to 50 to 60/sup 0/C. The radiation environment averages 20 to 50 mR/hr to the worker. Instruments and special equipment for data gathering at the sampler include pH, conductance, dissolved oxygen, dissolved hydrogen and nitrogen, hot leg and cold leg coolant temperatures, particle sizing, turbidimetry, filtration, and continuous strip chart recording.

  14. Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

    Energy Technology Data Exchange (ETDEWEB)

    Finn, P.A.

    1991-03-01

    Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

  15. Accurate representation of primary coolant system depressurization for high flux isotope reactor transients

    International Nuclear Information System (INIS)

    Wendel, M.W.

    1993-01-01

    Because the High Flux Isotope Reactor primary coolant system is typically all liquid, the pressure falls very rapidly if a leak forms in the primary piping. This depressurization is the predominate phenomenon in the consideration of the loss-of-coolant-accident analysis that has been completed recently for the High Flux Isotope Reactor Safety Analysis Report. Small differences in the rate of depressurization can significantly affect the safety margin. A RELAP5 thermal-hydraulic input model has been developed, but the capabilities of the RELAP5 code do not automatically take into account the effect on the system pressure of a stretching or shrinkage in the pressure boundary. Because this change in the pressure boundary is so important in an all-liquid system, a scheme has been developed to account for the effect implementation of the structural elasticity model involved using the control variable capability of RELAP5. During simulated transients involving rapid pressure changes, mass is added to or taken away from the primary system depending on whether the system is decreasing or increasing in pressure. By doing so, a more realistic system response to transients involving significant pressure changes is obtained. The model has been used to perform various HFIR transient simulations including loss-of-coolant accidents (LOCAs), loss of offsite power, and loss-of-secondary-cooling transients. Results for a small break LOCA are presented with and without the elasticity model in place

  16. Reducing forces during drilling brittle hard materials by using ultrasonic and variation of coolant

    Science.gov (United States)

    Schopf, C.; Rascher, R.

    2016-11-01

    The process of ultrasonic machining is especially used for brittle hard materials as the additional ultrasonic vibration of the tool at high frequencies and low amplitudes acts like a hammer on the surface. With this technology it is possible to drill holes with lower forces, therefor the machining can be done faster and the worktime is much less than conventionally. A three-axis dynamometer was used to measure the forces, which act between the tool and the sample part. A focus is set on the sharpness of the tool. The results of a test series are based on the Sauer Ultrasonic Grinding Centre. On the same machine it is possible to drill holes in the conventional way. Additional to the ultasonic Input the type an concentration of coolant is important for the Drilling-force. In the test there were three different coolant and three different concentrations tested. The combination of ultrasonic vibration and the right coolant and concentration is the best way to reduce the Forces. Another positive effect is, that lower drilling-forces produce smaller chipping on the edge of the hole. The way to reduce the forces and chipping is the main issue of this paper.

  17. Temperature and velocity field of coolant at inlet to WWER-440 core - evaluation of experimental data

    International Nuclear Information System (INIS)

    Jirous, F.; Klik, F.; Janeba, B.; Daliba, J.; Delis, J.

    1989-01-01

    Experimentally determined were coolant temperature and velocity fields at the inlet of the WWER-440 reactor core. The accuracy estimate is presented of temperature measurements and the relation is given for determining the resulting measurement error. An estimate is also made of the accuracy of solution of the system of equations for determining coefficients B kn using the method of the least square fit. Coefficients B kn represent the relative contribution of the mass flow of the k-th fuel assembly from the n-th loop and allow the calculation of coolant temperatures at the inlet of the k-th fuel assembly, when coolant temperatures in loops at reactor inlet are known. A comparison is made of the results of measurements on a hydrodynamic model of a WWER-440 reactor with results of measurements made at unit 4 of the Dukovany nuclear power plant. Full agreement was found for 32 model measurements and 6 reactor measurements. It may be assumed that the results of other model measurements obtained for other operating variants will also apply for an actual reactor. Their applicability may, however, only be confirmed by repeating the experiment on other WWER-440 reactors. (Z.M.). 5 figs., 7 refs

  18. Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1977-01-01

    Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

  19. Nuclear-radiation-actuated valve. [Patent application; for increasing coolant flow to blanket

    Science.gov (United States)

    Christiansen, D.W.; Schively, D.P.

    1982-01-19

    The present invention relates to a breeder reactor blanket fuel assembly coolant system valve which increases coolant flow to the blanket fuel assembly to minimize long-term temperature increases caused by fission of fissile fuel created from fertile fuel through operation of the breeder reactor. The valve has a valve first part (such as a valve rod with piston) and a valve second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics. The valve's first part is positioned to receive nuclear radiation from the nuclear reactor's fuel region. The valve's second part is positioned so that its nuclear radiation induced swelling is different from that of the valve's first part. The valve's second part also is positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system.

  20. Application of three-signal regulator for stabilizing fast reactor coolant temperature under emergency conditions

    International Nuclear Information System (INIS)

    Afanas'ev, V.A.; Gryazev, V.M.; Efimov, V.N.; Plyutinskij, V.I.; Tyufyagin, A.N.

    1979-01-01

    A possibility is analyzed for regulating coolant temperature under cram conditions of fart reactors by using signals of inertial thermocouples installed at the reactor outlet to control the coolant flow rate. To improve the regulation quality additional signals on the coolant flow rate and neutron flux are used altogether. The study has been conducted for the BOR-60 reactor. A technique is described for optimization of adjustment parameters of the regulator. Results of transient calculations for power levels of 20 and 60 MW are presented. The control system was tested on the BOR-60 reactor. As the temperature regulator a device was used manufactured on the basis of the RPIB type regulator. The experiments have demonstrated quite satisfactory performance of the control system: temperature deviations at the core outlet did not exceed 17K, and at the reactor outlet - 5K, whereas without the control system - 115 and 25K respectively. High efficiency of the temperature control system enables one to introduce it in fast power reactors

  1. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

    2012-09-01

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  2. Measurement of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Sandalls, F.J.

    1978-03-01

    Sulphur is an important element in some food chains and the release of radioactive sulphur to the environment must be closely controlled if the chemical form is such that it is available or potentially available for entering food chains. The presence of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor warranted a study to assess the quantity and chemical form of the radioactive sulphur in order to estimate the magnitude of the potential environmental hazard which might arise from the release of coolant gas from Civil Advanced Gas-Cooled Reactors. A combination of gas chromatographic and radiochemical analyses revealed carbonyl sulphide to be the only sulphur-35 compound present in the coolant gas of the Windscale Reactor. The concentration of carbonyl sulphide was found to lie in the range 40 to 100 x 10 -9 parts by volume and the sulphur-35 specific activity was about 20 mCi per gramme. The analytical techniques are described in detail. The sulphur-35 appears to be derived from the sulphur and chlorine impurities in the graphite. A method for the preparation of carbonyl sulphide labelled with sulphur-35 is described. (author)

  3. Water quality control device and water quality control method for reactor primary coolant system

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

    1995-01-01

    The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

  4. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    International Nuclear Information System (INIS)

    Chopra, O.K.; Shack, W.J.

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented

  5. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    International Nuclear Information System (INIS)

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

    2011-01-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  6. Distribution and behavior of tritium in the Coolant-Salt Technology Facility

    International Nuclear Information System (INIS)

    Mays, G.T.; Smith, A.N.; Engel, J.R.

    1977-04-01

    A 1000-MW(e) Molten-Salt Breeder Reactor (MSBR) is expected to produce 2420 Ci/day of tritium. As much as 60 percent of the tritium produced may be transported to the reactor steam system (assuming no retention by the secondary coolant salt), where it would be released to the environment. Such a release rate would be unacceptable. Experiments were conducted in an engineering-scale facility--the Coolant-Salt Technology Facility (CSTF)--to examine the potential of sodium fluoroborate, the proposed coolant salt for an MSBR, for sequestering tritium. The salt was believed to contain chemical species capable of trapping tritium. A series of 5 experiments--3 transient and 2 steady-state experiments--was conducted from July of 1975 through June of 1976 where tritium was added to the CSTF. The CSTF circulated sodium fluoroborate at temperatures and pressures typical of MSBR operating conditions. Results from the experiments indicated that over 90 percent of tritium added at steady-state conditions was trapped by sodium fluoroborate and appeared in the off-gas system in a chemically combined (water-soluble) form and that a total of approximately 98 percent of the tritium added at steady-state conditions was removed through the off-gas system overall

  7. Debris impact on emergency coolant recirculation - summary and conclusions

    International Nuclear Information System (INIS)

    Jain, Bhagwat; Hsia, Anthony; Armand, Yves; Mattei, Jean-Marie; Hyvaerinen, Juhani; Maqua, Michael; Puetter, Bernhard; Sandervaag, Oddbjoern; Vandewalle, Andre; Tombuyses, Beatrice; Pyy, Pekka; Royen, Jacques

    2004-01-01

    protection. - Amount of latent debris (e.g. use of qualified coatings, size of unprotected ferritic surfaces, cleanliness regime after outages). - Chemical effects due to NaOH. c. Sump strainer clogging may substantially increase the total core damage frequency. This statement depends strongly on the design features mentioned above and the assumption made to estimate the amount of insulation material reaching the sump strainer and the resulting pressure loss. Presently, different assumptions are used in different countries regarding debris generation and transport. d. Timely resolution of the sump strainer clogging issue is essential. Some participants presented solutions to the problem based on: - new strainer designs; - reduction of insulation material generation; - development of strainer cleaning procedures e.g. back-flushing. e. Timely solutions to the sump strainer clogging problem are plant design specific and debris specific. f. Assessment methods should continue to be enhanced. However, some countries have developed solutions or implemented compensatory measures. g. Chemical effects have to be taken into account. h. Latent debris continues to be of concern. i. Thin bed effect is one of the major concerns impacting strainer head loss and core cooling

  8. Material effect in the nuclear fuel-coolant interaction: Analyses of prototypic melt fragmentation and solidification in the KROTOS facility

    Czech Academy of Sciences Publication Activity Database

    Tyrpekl, V.; Piluso, P.; Bakardjieva, Snejana; Dugne, O.

    2014-01-01

    Roč. 186, č. 2 (2014), s. 229-240 ISSN 0029-5450 Institutional support: RVO:61388980 Keywords : fuel-coolant interaction * melt fragmentation * KROTOS facility Subject RIV: CA - Inorganic Chemistry Impact factor: 0.725, year: 2014

  9. A system for the discharge of gas bubbles from the coolant flow of a nuclear reactor cooled by forced circulation

    International Nuclear Information System (INIS)

    Markfort, D.; Kaiser, A.; Dohmen, A.

    1975-01-01

    In a reactor cooled by forced circulation the gas bubbles carried along with the coolant flow are separated before entering the reactor core or forced away into the external zones. For this purpose the coolant is radially guided into a plenum below the core and deflected to a tangential direction by means of flow guide elements. The flow runs spirally downwards. On the bubbles, during their dwell time in this channel, the buoyant force and a force towards the axis of symmetry of the tank are exerted. The major part of the coolant is directed into a radial direction by means of a guiding apparatus in the lower section of the channel and guided through a chimney in the plenum to the center of the reactor core. This inner chimney is enclosed by an outer chimney for the core edge zones through which coolant with a small share of bubbles is taken away. (RW) [de

  10. Bi-Metallic Additive Manufacturing Close-Out of Coolant Channels for Large Liquid Rocket Engine (LRE) Nozzles, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This NASA sponsored STTR project will investigate methods for close-out of large, liquid rocket engine nozzle, coolant channels utilizing robotic laser and...

  11. Dependence of model-simulated response to ozone depletion on stratospheric polar vortex climatology

    Science.gov (United States)

    Lin, Pu; Paynter, David; Polvani, Lorenzo; Correa, Gustavo J. P.; Ming, Yi; Ramaswamy, V.

    2017-06-01

    We contrast the responses to ozone depletion in two climate models: Community Atmospheric Model version 3 (CAM3) and Geophysical Fuild Dynamics Laboratory (GFDL) AM3. Although both models are forced with identical ozone concentration changes, the stratospheric cooling simulated in CAM3 is 30% stronger than in AM3 in annual mean, and twice as strong in December. We find that this difference originates from the dynamical response to ozone depletion, and its strength can be linked to the timing of the climatological springtime polar vortex breakdown. This mechanism is further supported by a variant of the AM3 simulation in which the southern stratospheric zonal wind climatology is nudged to be CAM3-like. Given that the delayed breakdown of the southern polar vortex is a common bias among many climate models, previous model-based assessments of the forced responses to ozone depletion may have been somewhat overestimated.

  12. Neutrino oscillations in strong magnetic fields

    International Nuclear Information System (INIS)

    Likhachev, G.G.; Studenikin, A.I.

    1994-07-01

    Neutrino conversion processes between two neutrino species and the corresponding oscillations induced by strong magnetic fields are considered. The value of the critical strength of magnetic field B cr as a function of characteristics of neutrinos in vacuum (Δm 2 ν , mixing angle θ), effective particle density of matter n eff , neutrino (transition) magnetic moment μ-tilde and energy E is introduced. It is shown that the neutrino conversion and oscillations effects induced by magnetic fields B ≥ B cr are important and may result in the depletion of the initial type of ν's in the bunch. A possible increase of these effects in the case when neutrinos pass through a sudden decrease of density of matter (''cross-boundary effect'') and applications to neutrinos from neutron stars and supernova are discussed. (author). 25 refs

  13. Experimental simulation of low rate primary coolant leaks. For the case of vessel head penetrations affected by through wall cracking

    International Nuclear Information System (INIS)

    You, D.; Feron, D.; Turluer, G.

    2002-01-01

    An experimental simulation of primary coolant leaks was carried out to determine how the composition of the leaking liquid would change. The experiment used the EVA experimental setup, specially designed for quantitatively investigating concentration phenomena driven by evaporation. The test showed that the final composition, obtained from a solution representative of the primary coolant at the beginning of the cycle, is highly concentrated and slightly acid. The experimental results are compared with those obtained using the MULTEQ software. (authors)

  14. Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

    International Nuclear Information System (INIS)

    Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

    2015-01-01

    At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

  15. A study on the sensitivity depletion laws for rhodium self-powered neutron detectors

    International Nuclear Information System (INIS)

    Kim, Gil Gon

    1999-02-01

    The rhodium self-powered neutron detectors (SPND) in a reactor core provide the operator with the on-line 3-dimensional nuclear power distribution. The signal produced by rhodium SPND is interpreted into the local neutron flux by using a sensitivity depletion law and the local neutron flux is interpreted into the local power by using a power conversion factor. This work on the sensitivity depletion laws for rhodium self-powered neutron detectors (SPND) is performed to improve the uncertainty of the sensitivity depletion law used in ABB-CE reactors employing a rhodium SPND and to develop a calculational tool for providing the sensitivity depletion laws to interpret the signal of the newly designed rhodium SPND into the local neutron flux. The calculational tools for a time dependent neutron flux distribution in the rhodium emitter during depletion and for a time dependent beta escape probability that a beta generated in the emitter is escaped into the collector were developed. Due to the cost, the exposure to the radiation, and the longer fuel cycle, there is a strong incentive that the loading density of an in-core instrumentation is reduced and the lifetime of the detector is lengthened. These objectives can be achieved by reducing the uncertainty which is amplified as it depletes. The calculational tools above provide the sensitivity depletion law and show the reduction of the uncertainty to about 1 % in interpreting the signal into the local neutron flux compared to the method employed by ABB-CE. The reduction in the uncertainty of 1 % in interpreting the signal into the local neutron flux is equivalent to the reduction in the uncertainty of 1 % or more in interpreting the signal into the local power and to the extension of the lifetime of rhodium SPND to about 10 % as reported by ABB-CE

  16. John Strong (1941 - 2006)

    CERN Multimedia

    Wickens, F

    Our friend and colleague John Strong was cruelly taken from us by a brain tumour on Monday 31st July, a few days before his 65th birthday John started his career working with a group from Westfield College, under the leadership of Ted Bellamy. He obtained his PhD and spent the early part of his career on experiments at Rutherford Appleton Laboratory (RAL), but after the early 1970s his research was focussed on experiments in CERN. Over the years he made a number of notable contributions to experiments in CERN: The Omega spectrometer adopted a system John had originally developed for experiments at RAL using vidicon cameras to record the sparks in the spark chambers; He contributed to the success of NA1 and NA7, where he became heavily involved in the electronic trigger systems; He was responsible for the second level trigger system for the ALEPH detector and spent five years leading a team that designed and built the system, which ran for twelve years with only minor interventions. Following ALEPH he tur...

  17. Stirring Strongly Coupled Plasma

    CERN Document Server

    Fadafan, Kazem Bitaghsir; Rajagopal, Krishna; Wiedemann, Urs Achim

    2009-01-01

    We determine the energy it takes to move a test quark along a circle of radius L with angular frequency w through the strongly coupled plasma of N=4 supersymmetric Yang-Mills (SYM) theory. We find that for most values of L and w the energy deposited by stirring the plasma in this way is governed either by the drag force acting on a test quark moving through the plasma in a straight line with speed v=Lw or by the energy radiated by a quark in circular motion in the absence of any plasma, whichever is larger. There is a continuous crossover from the drag-dominated regime to the radiation-dominated regime. In the crossover regime we find evidence for significant destructive interference between energy loss due to drag and that due to radiation as if in vacuum. The rotating quark thus serves as a model system in which the relative strength of, and interplay between, two different mechanisms of parton energy loss is accessible via a controlled classical gravity calculation. We close by speculating on the implicati...

  18. Strong-interaction nonuniversality

    International Nuclear Information System (INIS)

    Volkas, R.R.; Foot, R.; He, X.; Joshi, G.C.

    1989-01-01

    The universal QCD color theory is extended to an SU(3) 1 direct product SU(3) 2 direct product SU(3) 3 gauge theory, where quarks of the ith generation transform as triplets under SU(3)/sub i/ and singlets under the other two factors. The usual color group is then identified with the diagonal subgroup, which remains exact after symmetry breaking. The gauge bosons associated with the 16 broken generators then form two massive octets under ordinary color. The interactions between quarks and these heavy gluonlike particles are explicitly nonuniversal and thus an exploration of their physical implications allows us to shed light on the fundamental issue of strong-interaction universality. Nonuniversality and weak flavor mixing are shown to generate heavy-gluon-induced flavor-changing neutral currents. The phenomenology of these processes is studied, as they provide the major experimental constraint on the extended theory. Three symmetry-breaking scenarios are presented. The first has color breaking occurring at the weak scale, while the second and third divorce the two scales. The third model has the interesting feature of radiatively induced off-diagonal Kobayashi-Maskawa matrix elements

  19. Effect of cross-flow direction of coolant on film cooling effectiveness with one inlet and double outlet hole injection

    Directory of Open Access Journals (Sweden)

    Guangchao Li

    2012-12-01

    Full Text Available In order to study the effect of cross-flow directions of an internal coolant on film cooling performance, the discharge coefficients and film cooling effectiveness with one inlet and double outlet hole injections were simulated. The numerical results show that two different cross-flow directions of the coolant cause the same decrease in the discharge coefficients as that in the case of supplying coolant by a plenum. The different proportion of the mass flow out of the two outlets of the film hole results in different values of the film cooling effectiveness for three different cases of coolant supplies. The film cooling effectiveness is the highest for the case of supplying coolant by the plenum. At a lower blowing ratio of 1.0, the film cooling effectiveness with coolant injection from the right entrance of the passage is higher than that from the left entrance of the passage. At a higher blowing ratio of 2.0, the opposite result is found.

  20. Use of a multi-attribute utility theory for evaluating the best coolant material in transmutation reactors

    International Nuclear Information System (INIS)

    Yu, Dong Han; Han, Suk Joong; Kim, Do Hyung; Park, Won Suk

    1998-12-01

    In order to develop and design a good transmutation system, it is necessary first to select the best available coolant material for a reactor coolant system. Choosing the best coolant material may not be easy since there are several criteria associated with thermal performance, safety problem, cost problem, neutronic aspects. etc. The best option should be chosen based on the maximization of our needs in this situation. It is a challenging task. Decision theory can be employed to solve this type of problem. This report presents the feasibility study for evaluating the best coolant material in transmutation reactors based on the multi=attribute utility theory. The main problem presented here is how to logically evaluate candidate coolant materials under multiple criteria such as thermal performance, safety problem, cost problem, cost problem, neutronic aspects, etc. Since the current problem involves multiple criteria or attributes, first of all, the multi-attribute utility theory (MAUT) such as SMART and AHP has been extensively reviewed. Then, many candidate coolant material for transmutation reactors have been identified. The next step is to construct a value tree that express to reflect the relative importance of the attributes for overall evaluation. Finally, given these assignments, the final goal were obtained by manipulating these ranks through the value tree. The proposed approach is intended to help people be rational and logical in making decisions such complex tasks. (author). 8 refs., 7 tabs., 22 figs