WorldWideScience

Sample records for heat removal techniques

  1. Application of optimal estimation techniques to FFTF decay heat removal analysis

    International Nuclear Information System (INIS)

    Nutt, W.T.; Additon, S.L.; Parziale, E.A.

    1979-01-01

    The verification and adjustment of plant models for decay heat removal analysis using a mix of engineering judgment and formal techniques from control theory are discussed. The formal techniques facilitate dealing with typical test data which are noisy, redundant and do not measure all of the plant model state variables directly. Two pretest examples are presented. 5 refs

  2. After-heat removing device

    International Nuclear Information System (INIS)

    Iwashige, Kengo; Otsuka, Masaya; Yokoyama, Iwao; Yamakawa, Masanori.

    1990-01-01

    The present invention concerns an after-heat removing device for first reactors. A heat accumulation portion provided in a cooling channel of an after-heat removing device is disposed before a coil-like heat conduction pipe for cooling of the after-heat removing device. During normal reactor operation, the temperature in the heat accumulation portion is near the temperature of the high temperature plenum due to heat conduction and heat transfer from the high temperature plenum. When the reactor is shutdown and the after-heat removing device is started, coolants cooled in the air cooler start circulation. The coolants arriving at the heat accumulation portion deprive heat from the heat accumulation portion and, ion turn, increase their temperature and then reach the cooling coil. Subsequently, the heat calorie possessed in the heat accumulation portion is reduced and the after-heat removing device is started for the operation at a full power. This can reduce the thermal shocks applied to the cooling coil or structures in a reactor vessel upon starting the after-heat removing device. (I.N.)

  3. CRBRP decay heat removal systems

    International Nuclear Information System (INIS)

    Hottel, R.E.; Louison, R.; Boardman, C.E.; Kiley, M.J.

    1977-01-01

    The Decay Heat Removal Systems for the Clinch River Breeder Reactor Plant (CRBRP) are designed to adequately remove sensible and decay heat from the reactor following normal shutdown, operational occurrences, and postulated accidents on both a short term and a long term basis. The Decay Heat Removal Systems are composed of the Main Heat Transport System, the Main Condenser and Feedwater System, the Steam Generator Auxiliary Heat Removal System (SGAHRS), and the Direct Heat Removal Service (DHRS). The overall design of the CRBRP Decay Heat Removal Systems and the operation under normal and off-normal conditions is examined. The redundancies of the system design, such as the four decay heat removal paths, the emergency diesel power supplies, and the auxiliary feedwater pumps, and the diversities of the design such as forced circulation/natural circulation and AC Power/DC Power are presented. In addition to overall design and system capabilities, the detailed designs for the Protected Air Cooled Condensers (PACC) and the Air Blast Heat Exchangers (ABHX) are presented

  4. Innovative techniques for removing concrete surfaces

    International Nuclear Information System (INIS)

    McFarland, J.M.

    1980-01-01

    This report centers on the use of heat to decompose contaminated concrete to facilitate its removal. It discusses the use of electrical resistance heating and induction heating to cause differential expansion between the reinforcing steel and the concrete in order to spall the concrete. It introduces the concept of using induction heating to both decompose and spall steel impregnated concrete, acknowledging the work of Charles H. Henager in this field. The techniques are offered as theoretical and untested possibilities. Their practical application depends upon the effectiveness of alternatives and upon further development of these concepts

  5. Tritium removal by CO2 laser heating

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Mueller, D.

    1997-01-01

    Efficient techniques for rapid tritium removal will be necessary for ITER to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO 2 or Nd:Yag laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm 2 flux with an exposure time of order 10 ms is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally

  6. Tritium removal by CO2 laser heating

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Mueller, D.

    1997-10-01

    Efficient techniques for rapid tritium removal will be necessary for ITER (International Thermonuclear Experimental Reactor) to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO 2 or Nd:YAG laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm 2 flux with an exposure time of order 10 msec is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally

  7. Nuclear reactor auxiliary heat removal system

    International Nuclear Information System (INIS)

    Thompson, R.E.; Pierce, B.L.

    1977-01-01

    An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

  8. Application of heat pipes in nuclear reactors for passive heat removal

    Energy Technology Data Exchange (ETDEWEB)

    Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

  9. Study on diverse passive decay heat removal approach

    International Nuclear Information System (INIS)

    Lin Qian; Si Shengyi

    2012-01-01

    One of the most important principles for nuclear safety is the decay heat removal in accidents. Passive decay heat removal systems are extremely helpful to enhance the safety. In currently design of many advanced nuclear reactors, kinds of passive systems are proposed or developed, such as the passive residual heat removal system, passive injection system, passive containment cooling system. These systems provide entire passive heat removal paths from core to ultimate heat sink. Various kinds of passive systems for decay heat removal are summarized; their common features or differences on heat removal paths and design principle are analyzed. It is found that, these passive decay heat removal paths are similarly common on and connected by several basic heat transfer modes and steps. By the combinations or connections of basic modes and steps, new passive decay heat removal approach or diverse system can be proposed. (authors)

  10. Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol

    2014-01-01

    The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

  11. Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-10-15

    The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

  12. PWR passive plant heat removal assessment: Joint EPRI-CRIEPI advanced LWR studies

    International Nuclear Information System (INIS)

    1991-03-01

    An independent assessment of the capabilities of the PWR passive plant heat removal systems was performed, covering the Passive Residual Heat Removal (PRHR) System, the Passive Safety Injection System (PSIS) and the Passive Containment Cooling System (PCCS) used in a 600 MWe passive plant (e.g., AP600). Additional effort included a review of the test programs which support the design and analysis of the systems, an assessment of the licensability of the plant with regard to heat removal adequacy, and an evaluation of the use of the passive systems with a larger plant. The major conclusions are as follows. The PRHR can remove core decay heat, prevents the pressurizer from filling with water for a loss-of-feedwater transient, and provides safety-grade means for maintaining the reactor coolant system in a safe shutdown condition for the case where the non-safety residual heat removal system becomes unavailable. The PSIS is effective in maintaining the core covered with water for loss-of-coolant accident pipe breaks to eight inches. The PCCS has sufficient heat removal capability to maintain the containment pressure within acceptable limits. The tests performed and planned are adequate to confirm the feasibility of the passive heat removal system designs and to provide a database for verification of the analytical techniques used for the plant evaluations. Each heat removal system can perform in accordance with Regulatory requirements, with the exception that the PRHR system is unable to achieve the required cold shutdown temperature of 200 F within the required 36-hour period. The passive heat removal systems to be used for the 600 MWe plant could be scaled up to a 900 MWe passive plant in a straightforward manner and only minimal, additional confirmatory testing would be required. Sections have been indexed separately for inclusion on the data base

  13. Position paper -- Waste storage tank heat removal

    International Nuclear Information System (INIS)

    Stine, M.D.

    1995-01-01

    The purpose of this paper is to develop and document a position on the heat removal system to be used on the waste storage tanks currently being designed for the Multi-Function Waste Tank Facility (MWTF), project W-236A. The current preliminary design for the waste storage primary tank heat removal system consists of the following subsystems: (1) a once-through dome space ventilation system; (2) a recirculation dome space ventilation system; and (3) an annulus ventilation system. Recently completed and ongoing studies have evaluated alternative heat removal systems in an attempt to reduce system costs and to optimize heat removal capabilities. In addition, a thermal/heat transfer analysis is being performed that will provide assurance that the heat removal systems selected will be capable of removing the total primary tank design heat load of 1.25 MBtu/hr at an allowable operating temperature of 190 F. Although 200 F is the design temperature limit, 190 F has been selected as the maximum allowable operating temperature limit based on instrumentation sensitivity, instrumentation location sensitivity, and other factors. Seven options are discussed and recommendations are made

  14. Study on diverse passive decay heat removal approach and principle

    International Nuclear Information System (INIS)

    Lin Qian; Si Shengyi

    2012-01-01

    Decay heat removal in post-accident is one of the most important aspects concerned in the reactor safety analysis. Passive decay heat removal approach is used to enhance nuclear safety. In advanced reactors, decay heat is removed by multiple passive heat removal paths through core to ultimate heat sink by passive residual heat removal system, passive injection system, passive containment cooling system and so on. Various passive decay heat removal approaches are summarized in this paper, the common features and differences of their heat removal paths are analyzed, and the design principle of passive systems for decay heat removal is discussed. It is found that. these decay heat removal paths is combined by some basic heat transfer processes, by the combination of these basic processes, diverse passive decay heat removal approach or system design scheme can be drawn. (authors)

  15. Passive heat removal from containment

    International Nuclear Information System (INIS)

    Gou, P.F.; Townsend, H.E.

    1990-01-01

    This patent describes a heat removal system for removing heat from a containment of a nuclear reactor. It comprises: a sealed suppression chamber in the containment; means for venting steam from the nuclear reactor into the suppression chamber upon occurrence of an event requiring dissipation of heat from the nuclear reactor. The suppression chamber containing a quantity of water; the suppression chamber having a gas-containing space above the water; a heat exchanger disposed within the gas-containing space of the suppression chamber; the heat exchanger including an enclosed structure for holding a heat-exchange fluid; means for metering a supply of heat-exchange fluid to the heat exchanger to maintain a predetermined level thereof in the enclosed structure. The heat-exchange fluid boiling in the heat exchanger in consequence of heat transfer thereto from steam present in the suppression chamber; means for separating a heat-exchange fluid vapor in the heat exchanger from the heat-exchange fluid; and means for discharging the vapor immediately following its separation from heat-exchange fluid directly from the heat exchanger to a location exterior of the containment, whereby heat is discharged from the suppression chamber, and the containment is maintained at a temperature and pressure below its design value

  16. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Ohashi, Yukio.

    1990-01-01

    The after-heat removing system of the present invention removes the after heat generated in a reactor core without using dynamic equipments such as pumps or blowers. There are disposed a first heat exchanger for heating a heat medium by the heat in a reactor container and a second heat exchanger situated above the first heat exchanger for spontaneously air-cooling the heat medium. Recycling pipeways connect the first and the second heat exchangers to form a recycling path for the heat medium. Then, since the second heat exchanger for spontaneously air-cooling the heat medium is disposed above the first heat exchanger and they are connected by the recycling pipeways, the heat medium can be circulated spontaneously. Accordingly, dynamic equipments such as pumps or blowers are no more necessary. As a result, the after-heat removing system of the FBR type reactor of excellent safety and reliability can be obtained. (I.S.)

  17. Evaluation of Heat Removal Performance of Passive Decay Heat Removal system for S-CO{sub 2} Cooled Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The modular systems is able to be transported by large trailer. Moreover, dry cooling system is applied for waste heat removal. The characteristics of MMR takes wide range of construction area from coast to desert, isolated area and disaster area. In MMR, Passive decay heat removal system (PDHRS) is necessary for taking the advantage on selection of construction area where external support cannot be offered. The PDHRS guarantees to protect MMR without external support. In this research, PDHRS of MMR is introduced and decay heat removal performance is analyzed. The PDHRS guarantees integrity of reactor coolant system. The high level of decay heat (2 MW) can be removed by PDHRS without offsite power.

  18. After-heat removal system

    International Nuclear Information System (INIS)

    Yamamoto, Michiyoshi; Mitani, Shinji.

    1982-01-01

    Purpose: To prevent contamination of suppression pool water and intrusion of corrosion products into a nuclear reactor. Constitution: Upon stop of an after-heat removing system, reactor water contained in pipelines is drained out to a radioactive wastes processing facility at the time the cooling operation mode has been completed. At the same time, water is injected from a pure water supply system to the after-heat removing system to discharge corrosion product and activated materials while cleaning the inside of the pipelines. Then, pure water is held in the pipelines and it is discharged again and replaced with pure water before entering the cooling mode operation. Thereafter, the cooling mode operation upon reactor shutdown is performed. (Yoshino, Y.)

  19. Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen

    2005-08-29

    This report describes a three-year project to investigate the major factors of horizontal heat exchanger performance in passive containment heat removal from a light water reactor following a design basis accident LOCA (Loss of Coolant Accident). The heat exchanger studied in this work may be used in advanced and innovative reactors, in which passive heat removal systems are adopted to improve safety and reliability The application of horizontal tube-bundle condensers to passive containment heat removal is new. In order to show the feasibility of horizontal heat exchangers for passive containment cooling, the following aspects were investigated: 1. the condensation heat transfer characteristics when the incoming fluid contains noncondensable gases 2. the effectiveness of condensate draining in the horizontal orientation 3. the conditions that may lead to unstable condenser operation or highly degraded performance 4. multi-tube behavior with the associated secondary-side effects This project consisted of two experimental investigations and analytical model development for incorporation into industry safety codes such as TRAC and RELAP. A physical understanding of the flow and heat transfer phenomena was obtained and reflected in the analysis models. Two gradute students (one funded by the program) and seven undergraduate students obtained research experience as a part of this program.

  20. Assessment of the effectiveness of two heat removal techniques for permafrost protection

    DEFF Research Database (Denmark)

    Jørgensen, Anders Stuhr; Doré, Guy; Voyer, Érika

    2008-01-01

    Two mitigation techniques, an air convection embankment and an embankment of a granular material with an integrated heat drain, have been tested for the implementation in the shoulders of road and airfield embankments in permafrost regions. Both techniques will allow cold air to penetrate...... and calibrated on the SSE to verify the effects on the thermal regime of full-scale embankments. The results have shown that both techniques will cause a decrease in temperature, which will minimize or even possibly avoid permafrost degradation underneath the embankments. The laboratory results have also shown...

  1. Analysis of decay heat removal following loss of RHR

    International Nuclear Information System (INIS)

    Naff, S.A.; Ward, L.W.

    1991-01-01

    Recent plant experience has included many events occurring during outages at pressurized water reactors. A recent example is the loss of residual heat removal system event that occurred March 20, 1990 at the Vogtle-1 plant following refueling. Plant conditions during outages differ markedly from those prevailing at normal full-power operation on which most past research has concentrated. Specifically, during outages the core power is low, the coolant system may be in a drained state with air or nitrogen present, and various reactor coolant system closures may be unsecured. With the residual heat removal system operating, the core decay heat is readily removed. However, if the residual heat removal system capability is lost and alternative heat removal means cannot be established, heat up of the coolant could lead to core coolant boil-off, fuel rod heat up, and core damage. A study was undertaken by the Nuclear Regulatory Commission to identify what information was needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that might be used, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain into the reactor coolant system, core water boil-off, and reflux condensation cooling processes

  2. Prediction of Heat Removal Capacity of Horizontal Condensation Heat Exchanger submerged in Pool

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech., Yongin (Korea, Republic of); Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Park, Goon-Cherl [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    As representative passive safety systems, there are the passive containment cooling system (PCCS) of ESBWR, the emergency condenser system (ECS) of the SWR-1000, the passive auxiliary feed-water system (PAFS) of the APR+ and etc. During the nuclear power plant accidents, these passive safety systems can cool the nuclear system effectively via the heat transfer through the steam condensation, and then mitigate the accidents. For the optimum design and the safety analysis of the passive safety system, it is essential to predict the heat removal capacity of the heat exchanger well. The heat removal capacity of the horizontal condensation heat exchanger submerged in a pool is determined by a combination of a horizontal in-tube condensation heat transfer and a boiling heat transfer on the horizontal tube. Since most correlations proposed in the previous nuclear engineering field were developed for the vertical tube, there is a certain limit to apply these correlations to the horizontal tube. Therefore, this study developed the heat transfer model for the horizontal Ushaped condensation heat exchanger submerged in a pool to predict well the horizontal in-tube condensation heat transfer, the boiling heat transfer on the horizontal tube and the overall heat removal capacity of the heat exchanger using the best-estimate system analysis code, MARS.

  3. Tritium Removal by Laser Heating and Its Application to Tokamaks

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Guttadora, G.; Carpe, A.; Langish, S.; Young, K.M.; Nishi, M.; Shu, W.

    2001-01-01

    A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm 2 , and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed

  4. After-heat removal system of fast reactor

    International Nuclear Information System (INIS)

    Otsuka, Masaya; Shibata, Yoji; Ikeda, Takashi; Iwashige, Kengo; Yoneda, Yoshiyuki.

    1988-01-01

    Purpose: To remove after-heat by natural convection without disposing a movable portion even in a large-scaled reactor. Constitution: The exit of a reactor wall air-cooling duct disposed to the outside of a safety vessel is connected to the secondary inlet of an air cooler that conducts heat exchange with sodium in a high temperature plenum. That is, after-heat is removed only through the natural convection by a structure in which the reactor wall air-cooling duct and the secondary side of the air cooler are connected in series. Air exhausted from the exit of the air-cooling duct by the air cooler is further heated with sodium in the high temperature plenum. The flow rate of air flowing through the air-cooling duct is increased as compared with the case where the air cooler is not present. Accordingly, the flow rate of air at low temperature flowing through the inlet of the air duct is increased to increase the heat conduction amount. In this way, after-heat can be removed only by means of natural convection without providing movable portions even in a large-scaled reactor with the thermal power in excess of 2,000 MW. (Horiuchi, T.)

  5. Advances in technologies for decay heat removal

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Berkovich, V.; Bianchi, A.; Chen B.; Meseth, J.; Vecchiarelli, J.; Vidard, M.

    1999-01-01

    The various decay heat removal concepts that have been used for the evolutionary water reactor plant designs developed worldwide are examined and common features identified. Although interesting new features of the 'classical' plants are mentioned, the emphasis is on passive core and containment decay heat removal systems. The various systems are classified according to the function they have to accomplish; they often share common characteristics and similar equipment. (author)

  6. Heat removing under hypersonic conditions

    Directory of Open Access Journals (Sweden)

    Semenov Mikhail E.

    2016-01-01

    Full Text Available In this paper we consider the heat transfer properties of the axially symmetric body with parabolic shape at hypersonic speeds (with a Mach number M > 5. We use the numerical methods based on the implicit difference scheme (Fedorenko method with direct method based on LU-decomposition and iterative method based on the Gauss-Seigel method. Our numerical results show that the heat removing process should be performed in accordance with the nonlinear law of heat distribution over the surface taking into account the hypersonic conditions of motion.

  7. Heat release, time required, and cleaning ability of MTwo R and ProTaper universal retreatment systems in the removal of filling material.

    Science.gov (United States)

    Bramante, Clovis Monteiro; Fidelis, Natasha Siqueira; Assumpção, Tatiana Santos; Bernardineli, Norberti; Garcia, Roberto Brandão; Bramante, Alexandre Silva; de Moraes, Ivaldo Gomes

    2010-11-01

    This ex vivo study evaluated the heat release, time required, and cleaning efficacy of MTwo (VDW, Munich, Germany) and ProTaper Universal Retreatment systems (Dentsply/Maillefer, Ballaigues, Switzerland) and hand instrumentation in the removal of filling material. Sixty single-rooted human teeth with a single straight canal were obturated with gutta-percha and zinc oxide and eugenol-based cement and randomly allocated to 3 groups (n = 20). After 30-day storage at 37 °C and 100% humidity, the root fillings were removed using ProTaper UR, MTwo R, or hand files. Heat release, time required, and cleaning efficacy data were analyzed statistically (analysis of variance and the Tukey test, α = 0.05). None of the techniques removed the root fillings completely. Filling material removal with ProTaper UR was faster but caused more heat release. Mtwo R produced less heat release than the other techniques but was the least efficient in removing gutta-percha/sealer. ProTaper UR and MTwo R caused the greatest and lowest temperature increase on root surface, respectively; regardless of the type of instrument, more heat was released in the cervical third. Pro Taper UR needed less time to remove fillings than MTwo R. All techniques left filling debris in the root canals. Copyright © 2010 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

  8. Solution of heat removal from nuclear reactors by natural convection

    Directory of Open Access Journals (Sweden)

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  9. Decay heat removal for the liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Zemanick, P.P.; Brown, N.W.

    1975-01-01

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. A statement of the high reliability of the Clinch River Breeder reactor Plant decay heat removal systems and a summary of the supporting arguments is presented. (U.S.)

  10. Decay Heat Removal for the Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zemanick, P. P.; Brown, N. W.

    1975-10-15

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. The paper closes with a statement of the high reliability of the Clinch River Breeder Reactor Plant decay heat removal systems and a summary of the supporting arguments. (author)

  11. Highly heat removing radiation shielding material

    International Nuclear Information System (INIS)

    Asano, Norio; Hozumi, Masahiro.

    1990-01-01

    Organic materials, inorganic materials or metals having excellent radiation shielding performance are impregnated into expanded metal materials, such as Al, Cu or Mg, having high heat conductivity. Further, the porosity of the expanded metals and combination of the expanded metals and the materials to be impregnated are changed depending on the purpose. Further, a plurality of shielding materials are impregnated into the expanded metal of the same kind, to constitute shielding materials. In such shielding materials, impregnated materials provide shielding performance against radiation rays such as neutrons and gamma rays, the expanded metals provide heat removing performance respectively and they act as shielding materials having heat removing performance as a whole. Accordingly, problems of non-informity and discontinuity in the prior art can be dissolved be provide materials having flexibility in view of fabrication work. (T.M.)

  12. An innovative pool with a passive heat removal system

    International Nuclear Information System (INIS)

    Vitale Di Maio, Damiano; Naviglio, Antonio; Giannetti, Fabio; Manni, Fabio

    2012-01-01

    Heat removal systems are of primary importance in several industrial processes. As heat sink, a water pool or atmospheric air may be selected. The first solution takes advantage of high heat transfer coefficient with water but it requires active systems to maintain a constant water level; the second solution takes benefit from the unlimited heat removal capacity by air, but it requires a larger heat exchanger to compensate the lower heat transfer coefficient. In NPPs (nuclear power plants) during a nuclear reactor shutdown, as well as in some chemical plants to control runaway reactions, it is possible to use an innovative heat sink that joins the advantages of the two previous solutions. This solution is based on a special heat exchanger submerged in a water pool designed so that when heat removal is requested, active systems are not required to maintain the water level; due to the special design, when the pool is empty, atmospheric air becomes the only heat sink. The special heat exchanger design allows to have a heat exchanger without being oversized and to have a system able to operate for unlimited period without external interventions. This innovative system provides an economic advantage as well as enhanced safety features.

  13. Study on thermal-hydraulic phenomena identification of passive heat removal facilities

    International Nuclear Information System (INIS)

    Park, J. Y.

    2011-01-01

    Recently, passive heat removal facilities have been integral features of new generation or future reactor designs worldwide. This is because the passive heat removal facilities depending on a natural force such as buoyancy can give much higher operational reliability compared to active heat removal facilities depending on pumped fluid flow and as a result they can decrease core damage frequency of a nuclear power plant drastically ever achievable before. Keeping pace with this global trend, SMART and APR+ reactors also have introduced passive heat removal features such as a passive residual heat removal system (PRHRS) and a passive auxiliary feed water system (PAFS) in their designs. Since many thermal-hydraulic (T-H) phenomena including steam condensation are involved during operation of the passive heat removal facilities, they ought to be properly simulated by T-H codes such as MARS-KS and RELAP5 in order to guarantee reliable safety analysis by these codes. Unfortunately, however, these T-H codes are not well validated with respect to phenomena related to passive heat removal mechanism because previous focus on these codes validation was mainly on the LB LOCA and resulting phenomena. To resolve this gap, Korea Institute of Nuclear Safety has initiated a research program on the development of safety analysis technology for passive heat removal facilities. The main target of this program is PRHRS and PAFS in SMART and APR+ reactors and through this program, validation of capability of existing T-H codes and improvement of codes regarding passive facilities analysis are to be sought. In part of this research, T-H phenomena important to passive heat removal facilities (PRHRS and PAFS) are investigated in the present study

  14. Cyclic process for producing methane from carbon monoxide with heat removal

    Science.gov (United States)

    Frost, Albert C.; Yang, Chang-lee

    1982-01-01

    Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

  15. Residual heat removal system diagnostic advisor

    International Nuclear Information System (INIS)

    Tripp, L.

    1991-01-01

    This paper reports on the Residual Heat Removal System (RHRS) Diagnostic Advisor which is an expert system designed to alert the operators to abnormal conditions that exits in the RHRS and offer advice about the cause of the abnormal conditions. The Advisor uses a combination of rule-based and model-based diagnostic techniques to perform its functions. This diagnostic approach leads to a deeper understanding of the RHRS by the Advisor and consequently makes it more robust to unexpected conditions. The main window of the interactive graphic display is a schematic diagram of the RHRS piping system. When a conclusion about a failed component can be reached, the operator can bring up windows that describe the failure mode of the component and a brief explanation about how the Advisor arrived at its conclusion

  16. Assessment of ASME code examinations on regenerative, letdown and residual heat removal heat exchangers

    International Nuclear Information System (INIS)

    Gosselin, Stephen R.; Cumblidge, Stephen E.; Anderson, Michael T.; Simonen, Fredric A.; Tinsley, G A.; Lydell, B.; Doctor, Steven R.

    2005-01-01

    Inservice inspection requirements for pressure retaining welds in the regenerative, letdown, and residual heat removal heat exchangers are prescribed in Section XI Articles IWB and IWC of the ASME Boiler and Pressure Vessel Code. Accordingly, volumetric and/or surface examinations are performed on heat exchanger shell, head, nozzle-to-head, and nozzle-to-shell welds. Inspection difficulties associated with the implementation of these Code-required examinations have forced operating nuclear power plants to seek relief from the U.S. Nuclear Regulatory Commission. The nature of these relief requests are generally concerned with metallurgical, geometry, accessibility, and radiation burden. Over 60% of licensee requests to the NRC identify significant radiation exposure burden as the principle reason for relief from the ASME Code examinations on regenerative heat exchangers. For the residual heat removal heat exchangers, 90% of the relief requests are associated with geometry and accessibility concerns. Pacific Northwest National Laboratory was funded by the NRC Office of Nuclear Regulatory Research to review current practice with regard to volumetric and/or surface examinations of shell welds of letdown heat exchangers regenerative heat exchangers and residual (decay) heat removal heat exchangers Design, operating, common preventative maintenance practices, and potential degradation mechanisms are reviewed. A detailed survey of domestic and international PWR-specific operating experience was performed to identify pressure boundary failures (or lack of failures) in each heat exchanger type and NSSS design. The service data survey was based on the PIPExp- database and covers PWR plants worldwide for the period 1970-2004. Finally a risk assessment of the current ASME Code inspection requirements for residual heat removal, letdown, and regenerative heat exchangers is performed. The results are then reviewed to discuss the examinations relative to plant safety and

  17. CAREM-25: Residual heat removal system

    International Nuclear Information System (INIS)

    Arvia, Roberto P.; Coppari, Norberto R.; Gomez de Soler, Susana M.; Ramilo, Lucia B.

    2000-01-01

    The objective of this work was the definition and consolidation of the residual heat removal system for the CAREM 25 reactor. The function of this system is cool down the primary circuit, removing the core decay heat from hot stand-by to cold shutdown and during refueling. In addition, this system heats the primary water from the cold shutdown condition to hot stand-by condition during the reactor start up previous to criticality. The system has been designed according to the requirements of the standards: ANSI/ANS 51.1 'Nuclear safety criteria for the design of stationary PWR plants'; ANSI/ANS 58.11 'Design criteria for safe shutdown following selected design basis events in light water reactors' and ANSI/ANS 58.9 'Single failure criteria for light water reactor safety-related fluid systems'. The suggested design fulfills the required functions and design criteria standards. (author)

  18. Techniques for removing contaminated concrete surfaces

    International Nuclear Information System (INIS)

    Halter, J.M.; Sullivan, R.G.

    1981-01-01

    This discussion compares various techniques that have been used to clean concrete surfaces by removing the surface. Three techniques which have been investigated by the Pacific Northwest Laboratory for removing surfaces are also described: the water cannon, the concrete spaller, and high-pressure water jet. The equipment was developed with the assumption that removal of the top 1/8 to 1/4 in. of surface would remove most of the contamination. If the contamination has gone into cracks or deep voids in the surface, the removal processes can be repeated until the surface is acceptable

  19. Study on decay heat removal capability of reactor vessel auxiliary cooling system

    International Nuclear Information System (INIS)

    Nishi, Y.; Kinoshita, I.

    1991-01-01

    The reactor vessel auxiliary cooling system (RVACS) is a simple, Passive decay heat removal system for an LMFBR. However, the heat removal capacity of this system is small compared to that of an immersed type of decay heat exchanger. In this study, a high-porosity porous body is proposed to enhance the RVACS's heat transfer performance to improve its applicability. The objectives of this study are to propose a new method which is able to use thermal radiation effectively, to confirm its heat removal capability and to estimate its applicability limit of RVACS for an LMFBR. Heat transfer tests were conducted in an experimental facility with a 3.5 m heat transfer height to evaluate the heat transfer performance of the high-porosity porous body. Using the experimental results, plant transient analyses were performed for a 300 MWe pool type LMFBR under a Total Black Out (TBO) condition to confirm the heat removal capability. Furthermore, the relationship between heat removal capability and thermal output of a reactor were evaluated using a simple parameter model

  20. Preliminary Analysis on Heat Removal Capacity of Passive Air-Water Combined Cooling Heat Exchanger Using MARS

    International Nuclear Information System (INIS)

    Kim, Seung-Sin; Jeon, Seong-Su; Hong, Soon-Joon; Bae, Sung-Won; Kwon, Tae-Soon

    2015-01-01

    Current design requirement for working time of PAFS heat exchanger is about 8 hours. Thus, it is not satisfied with the required cooling capability for the long term SBO(Station Black-Out) situation that is required to over 72 hours cooling. Therefore PAFS is needed to change of design for 72 hours cooling. In order to acquirement of long terms cooling using PAFS, heat exchanger tube has to be submerged in water tank for long time. However, water in the tank is evaporated by transferred heat from heat exchanger tubes, so water level is gradually lowered as time goes on. The heat removal capacity of air cooling heat exchanger is core parameter that is used for decision of applicability on passive air-water combined cooling system using PAFS in long term cooling. In this study, the development of MARS input model and plant accident analysis are performed for the prediction of the heat removal capacity of air cooling heat exchanger. From analysis result, it is known that inflow air velocity is the decisive factor of the heat removal capacity and predicted air velocity is lower than required air velocity. But present heat transfer model and predicted air velocity have uncertainty. So, if changed design of PAFS that has over 4.6 kW heat removal capacity in each tube, this type heat exchanger can be applied to long term cooling of the nuclear power plant

  1. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    Hart, R.S.

    1997-01-01

    CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D 2 O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

  2. Experience with after-shutdown decay heat removal - BWRs and PWRs

    International Nuclear Information System (INIS)

    Haugh, J.J.; Mollerus, F.J.; Booth, H.R.

    1992-01-01

    Boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) make use of residual heat removal systems (RHRSs) during reactor shutdown. RHRS operational events involving an actual loss or significant degradation of an RHRS during shutdown heat removal are often prompted or aggravated by complex, changing plant conditions and by concurrent maintenance operations. Events involving loss of coolant inventory, loss of decay heat removal capability, or inadvertent pressurization while in cold shutdown have occurred. Because fewer automatic protective fetures are operative during cold shutdowns, both prevention and termination of events depend heavily on operator action. The preservation of RHRS cooling should be an important priority in all shutdown operations, particularly where there is substantial decay heat and a reduced water inventory. 13 refs., 3 figs., 4 tabs

  3. Design and analysis of a new passive residual heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Xing [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Peng, Minjun, E-mail: heupmj@163.com [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Yuan, Xiao [Guangxi Fangchenggang Nuclear Power Co., Ltd (China); Xia, Genglei [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China)

    2016-07-15

    Highlights: • An air cooling passive residual heat removal System (PRHRs) is designed. • Using RELAP5/MOD3.4 code to analyze the operation characteristics of the PRHRs. • Noncondensable gas is used to simulate the hydrodynamic behavior in the air cooling tower. • The natural circulations could respectively establish in the primary circuit and the PRHRs circuit. • The PRHRs could remove the residual heat effectively. - Abstract: The inherent safety functions will mitigate the consequences of the accidents, and it can be accomplished through the passive safety systems which employed in the typical pressurized water reactor (PWR). In this paper, a new passive residual heat removal system (PRHRS) is designed for a typical nuclear power plant. PRHRS consists of a steam generator (SG), a cooling tank with two groups of cooling pipes, an air-cooling heat exchanger (AHX), an air-cooling tower, corresponding pipes and valves. The cooling tank which works as an intermediate buffer device is used to transfer the core decay heat to the AHX, and then the core decay heat will be removed to the atmosphere finally. The RELAP5/MOD3.4 code is used to analyze the operation characteristics of PRHRS and the primary loop system. It shows PRHRS could remove the decay heat from the primary loop effectively, and the natural circulations can be established in the primary circuit and the PRHRS circuit respectively. Furthermore, the sensitivity study has also been done to research the effect of various factors on the heat removal capacity.

  4. Passive Decay Heat Removal System for Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

  5. The kinetics of removal of heat-induced excess nuclear protein

    International Nuclear Information System (INIS)

    Roti, J.L.R.; Uygur, N.; Higashikubo, R.

    1984-01-01

    To investigate the role of protein content, temperature and heating time in the removal of heat-induced excess protein associated with the isolated nucleus, the kinetics of protein removal was monitored for 6 to 8 hours following exposure to 7 hyperthermic protocols. Four of these (47 0 C-7.5 min., 46 0 C-15 min., 45 0 C-30 min., and 44 0 C-60 min.) resulted in a nuclear protein content approximately twice that of nuclei from unheated cells (2.05 +- .14) following heat exposure. Three protocols (45 0 C-15 min., 44 0 C-30 min. and 43 0 C-60 min.) resulted in a nuclear protein content approximately 1.6 times normal (1.63 +- .12). If nuclear protein content were the only determinant in the recovery rate, then the same half time for nuclear protein removal would be expected within each group of protocols. Rate constants for nuclear protein removal were obtained by regression analysis. The half-time for nuclear protein removal increased with decreasing temperature and increasing heating time for the same nuclear protein content. This result suggests that the heating time and temperature are more of a determinant in the removal kinetics than protein content alone. Extended kinetics of recovery (to 36 hours) showed incomplete recovery and a secondary increase in protein associated with the isolated nucleus. These results were due to cell-cycle rearrangement (G/sub 2/ block) and unbalanced growth

  6. The heat engine cycle, the heat removal cycle, and ergonomics of the control room displays

    International Nuclear Information System (INIS)

    Beltracchi, L.

    1986-01-01

    This paper discusses and illustrates the ergonomics of an integrated display, which will allow operators to monitor the heat engine cycle during normal operation of the plant, and the heat removal cycle during emergency operation of the plant. A computer-based iconic display is discussed as an overview to monitor these cycles. Specific emphasis is placed upon the process variables and process functions within each cycle, and the action of control systems and engineered safeguard systems within each cycle. This paper contains examples of display formats for the heat engine cycle and the heat removal cycle in a pressurized water reactor

  7. Performance of the prism reactor's passive decay heat removal system

    International Nuclear Information System (INIS)

    Magee, P.M.; Hunsbedt, A.

    1989-01-01

    The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

  8. Passive heat removal characteristics of SMART

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jae Kwang; Kang, Hyung Seok; Yoon, Joo Hyun; Kim, Hwan Yeol; Cho, Bong Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRS of the SMART has excellent passive heat removal characteristics. 2 refs., 4 figs., 1 tab. (Author)

  9. Experimental study on heat pipe heat removal capacity for passive cooling of spent fuel pool

    International Nuclear Information System (INIS)

    Xiong, Zhenqin; Wang, Minglu; Gu, Hanyang; Ye, Cheng

    2015-01-01

    Highlights: • A passively cooling SFP heat pipe with an 8.2 m high evaporator was tested. • Heat removed by the heat pipe is in the range of 3.1–16.8 kW. • The heat transfer coefficient of the evaporator is 214–414 W/m 2 /K. • The heat pipe performance is sensitive to the hot water temperature. - Abstract: A loop-type heat pipe system uses natural flow with no electrically driven components. Therefore, such a system was proposed to passively cool spent fuel pools during accidents to improve nuclear power station safety especially for station blackouts such as those in Fukushima. The heat pipe used for a spent fuel pool is large due to the spent fuel pool size. An experimental heat pipe test loop was developed to estimate its heat removal capacity from the spent fuel pool during an accident. The 7.6 m high evaporator is heated by hot water flowing vertically down in an assistant tube with a 207-mm inner diameter. R134a was used as the potential heat pipe working fluid. The liquid R134a level was 3.6 m. The tests were performed for water velocities from 0.7 to 2.1 × 10 −2 m/s with water temperatures from 50 to 90 °C and air velocities from 0.5 m/s to 2.5 m/s. The results indicate significant heat is removed by the heat pipe under conditions that may occur in the spent fuel pool

  10. A passive decay-heat removal system for an ABWR based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2017-01-15

    Highlights: • A passive decay heat removal system for an ABWR is discussed using combined system of the reactor and an air cooler. • Effect of number of pass of the finned heat transfer tubes on heat removal is investigated. • The decay heat can be removed by air coolers with natural convection. • Two types of air cooler are evaluated, i.e., steam condensing and water cooling types. • Measures how to improve the heat removal rate and to make compact air cooler are discussed. - Abstract: This paper describes the capability of an air cooling system (ACS) operated under natural convection conditions to remove decay heat from the core of an Advanced Boiling Water Reactor (ABWR). The motivation of the present research is the Fukushima Severe Accident (SA). The plant suffered damages due to the tsunami and entered a state of Station Blackout (SBO) during which seawater cooling was not available. To prevent this kind of situation, we proposed a passive decay heat removal system (DHRS) in the previous study. The plant behavior during the SBO was calculated using the system code NETFLOW++ assuming an ABWR with the ACS. However, decay heat removal under an air natural convection was difficult. In the present study, a countermeasure to increase heat removal rate is proposed and plant transients with the ACS are calculated under natural convection conditions. The key issue is decreasing pressure drop over the tube banks in order to increase air flow rate. The results of the calculations indicate that the decay heat can be removed by the air natural convection after safety relief valves are actuated many times during a day. Duct height and heat transfer tube arrangement of the AC are discussed in order to design a compact and efficient AC for the natural convection mode. As a result, a 4-pass heat transfer tubes with 2-row staggered arrangement is the candidate of the AC for the DHRS under the air natural convection conditions. The heat removal rate is re-evaluated as

  11. A passive decay heat removal system for LWRs based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2015-05-15

    Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

  12. After-heat removing device in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mizuno, K [Nippon Atomic Industry Group Co. Ltd., Tokyo

    1977-01-14

    Purpose: To prevent water hammer in a BWR type reactor or the like by moving water in pipe lines having stagnant portions in an after-heat removing device. Constitution: To a reactor container, is provided a recycling pump which constitutes a closed loop type recycling system in a nuclear power plant together with a pressure vessel and pipe lines. A pump and a heat exchanger are provided outside of the reactor container and they are connected to up- and down-streams of the recycling pump to form an after-heat removing device in the plant. Upon shutdown of the nuclear power plant, since water in the stagnant portion flows to the intake port of the recycling pump and water from the reactor is spontaneously supplemented thereafter to the stagnant portion, neither pressurized water nor heated steam is generated and thus water hammer is prevented.

  13. Modes of heat removal from a heat-generating debris bed

    International Nuclear Information System (INIS)

    Squarer, D.; Hochreiter, L.E.; Piecznski, A.T.

    1984-01-01

    In the worst hypothetical accident in a light water reactor, when all protection systems fail, the core could be converted into a deep particulate bed either in-vessel or ex-vessel. The containment of such an accident depends on the coolability of a heat-generating debris bed. Some recent experimental and analytical studies that are concerned with heat removal from such a particulate bed are reviewed. Studies have indicated that bed dryout flux and, therefore, the heat removal rate from the particulate bed increases with the particle diameter (i.e., the permeability) for pool boiling conditions and can exceed the critical heat flux of a flat plate. Bed dryout in a large particle bed (i.e., a few millimetres) was found to be closely related to the ''flooding'' limit of the bed. Dryout under forced flow conditions was found to be affected by both forced and natural convection for mass flow rate smaller than m /SUB cr/ , whereas above this mass flow rate, bed dryout is proportional to the mass flow rate. Recent analyses were found to be in agreement with experimental data; however, additional research is needed to assess factors not accounted for in previous studies (e.g., effect of pressure, multidimensionality, stratification, etc.). Based on the expected pressure and particle sizes in a postulated severe accident sequence, a debris bed should be coolable, given a sufficient water supply

  14. Cyclic process for producing methane in a tubular reactor with effective heat removal

    Science.gov (United States)

    Frost, Albert C.; Yang, Chang-Lee

    1986-01-01

    Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

  15. Heat removal capability of core-catcher with inclined cooling channels

    International Nuclear Information System (INIS)

    Suzuki, Y.; Tahara, M.; Kurita, T.; Hamazaki, R.; Morooka, S.

    2009-01-01

    A core-catcher is one of the mitigation systems that provide functions of molten corium cooling and stabilization during a severe accident. Toshiba has been developing a compact core-catcher to be placed at the lower drywell floor in the containment vessel for the next generation BWR as well as near term ABWR. This paper presents the evaluation of heat removal capability of the core-catcher with inclined cooling channels, our verification status and plan. The heat removal capability of the core-catcher is analyzed by using the newly developed two-phase flow analysis code which incorporates drift flux parameters for inclined channels and the CHF correlation obtained from SULTAN tests. Effects of geometrical parameters such as the inclination and the gap size of the cooling channel on the heat removal capability are also evaluated. These results show that the core-catcher has sufficient capability to cool the molten corium during a severe accident. Based on the analysis, it has been shown that the core-catcher has an efficient capability of heat removal to cool the molten corium. (author)

  16. Postaccident heat removal. II. Heat transfer from an internally heated liquid to a melting solid

    International Nuclear Information System (INIS)

    Faw, R.E.; Baker, L. Jr.

    1976-01-01

    Microwave heating has been used in studies of heat transfer from a horizontal layer of internally heated liquid to a melting solid. Experiments were designed to simulate heat transfer and meltthrough processes of importance in the analysis of postaccident heat removal capabilities of nuclear reactors. Glycerin, heated by 2.45-GHz microwave radiation, was used to simulate molten fuel. Paraffin wax was used to simulate a melting barrier confining the fuel. Experimentally measured heat fluxes and melting rates were consistent with a model based on downward heat transfer by conduction through a stagnant liquid layer and upward heat transfer augmented by natural convection. Melting and displacement of the barrier material occurred by upward-moving droplets randomly distributed across the melting surface. Results indicated that the melting and displacement process had no effect on the heat transfer process

  17. Sensitivity analysis for maximum heat removal from debris in the lower head

    International Nuclear Information System (INIS)

    Kim, Yong Hoon; Suh, Kune Y.

    2000-01-01

    Sensitivity analyses were performed to determine the maximum heat removal capability from the debris and the reactor pressure vessel (RPV) wall through the gap that may be formed during a core melt relocation accident. Cases studied included four different nuclear power plant (TMI-2,KORI-2,YGN 3and4 and KNGR) per the thermal opower output. Results of the analysis show that the heat removal through gap cooling relative to flooding is efficacious as much as about 40% of the core material accumulated in the lower plenum in case of the TMI-2 reactor. In excess of 40%, however, the gap cooling alone was found not to be enough for heat removal from the core debris. There being uncertaainties aoboout the assumptions made in the present study,the analyses yield consistent results. If different cooling effects are considered, heat removal may be greatly enhanced. The LAVA experiements were performed at the Korea Atomic Energy Research Institute (KAERI) using al 2 O 3 /Fe thermite melt relocating down to the scaled vessel of a reactor lower head filled with preheated water. Test results indicated a cooling effect of water ingression through the debris-to-vessel gap and the intra-debris pores and crevices. If the cooling capacity of the intra-debris pores and crevices is comparable to debris-to-vessel heat removal capability, heat removal from the debris will be greatly augmented than heat removal by the gap cooling alone. The three nuclear reactor (KORI-2, YGN 3and4 and KNGR) calculation results for heat removal through the debris-to-vessel gap size of about 1mm were compared with the TMI-2 reactor calculation results for the case of gap cooling alone. (author)

  18. Passive heat removal system with injector-condenser

    Energy Technology Data Exchange (ETDEWEB)

    Soplenkov, K I [All-Russian Inst. of Nuclear Power Plant Operation, Electrogorsk Research and Engineering Centre of Nuclear Power Safety (Russian Federation)

    1996-12-01

    The system described in this paper is a passive system for decay heat removal from WWERs. It operates off the secondary side of the steam generators (SG). Steam is taken from the SG to operate a passive injector pump which causes secondary fluid to be pumped through a heat exchanger. Variants pass either water or steam from the SG through the heat exchanger. There is a passive initiation scheme. The programme for experimental and theoretical validation of the system is described. (author). 8 figs.

  19. Heat removal in INTOR via a toroidal limiter

    International Nuclear Information System (INIS)

    Mioduszewski, P.

    1981-01-01

    In the present paper the potential of removing about 100 MW of thermal plasma power via a toroidal limiter in INTOR is studied. The heat flux distributions on various limiter configurations are calculated and the thermal response of a graphite tile limiter is estimated on the base of a one-dimensional heat conduction approach. The evaporation rates which have to be expected for the given energy flux densities and radiation cooled graphite tiles are evaluated. According to the present understanding it should be possible to remove 100 MW power from the INTOR plasma via a radiation cooled toroidal limiter. (author)

  20. Decision Document for Heat Removal from High-Level Waste Tanks

    International Nuclear Information System (INIS)

    WILLIS, W.L.

    2000-01-01

    This document establishes the combination of design and operational configurations that will be used to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. The chosen method--to use the primary and annulus ventilation systems to remove heat from the high-level waste tanks--is documented herein

  1. Application study of the heat pipe to the passive decay heat removal system of the modular HTR

    International Nuclear Information System (INIS)

    Ohashi, K.; Okamoto, F.; Hayakawa, H.; Hayashi, T.

    2001-01-01

    To investigate the applicability of the heat pipe to the decay hat removal (DHR) system of the modular HTRs, preliminary study of the Heat Pipe DHR System was performed. The results show that the Heat Pipe DHR System is applicable to the modular HTRs and its heat removal capability is sufficient. Especially by applying the variable conductance heat pipe, the possibility of a fully passive DHR system with lower heat loss during normal operation is suggested. The experiments to obtain the fundamental characteristics data of the variable conductance heat pipe were carried out. The experimental results show very clear features of self-control characteristics. The experimental results and the experimental analysis results are also shown. (author)

  2. Nuclear reactor with makeup water assist from residual heat removal system

    Science.gov (United States)

    Corletti, Michael M.; Schulz, Terry L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  3. Nuclear reactor with makeup water assist from residual heat removal system

    International Nuclear Information System (INIS)

    Corletti, M.M.; Schulz, T.L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures

  4. Nuclear reactor with makeup water assist from residual heat removal system

    Science.gov (United States)

    Corletti, M.M.; Schulz, T.L.

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  5. Safety studies on heat transport and afterheat removal for GCR accident conditions

    International Nuclear Information System (INIS)

    Hishida, Makoto

    1996-01-01

    The IAEA coordinated an international research program on 'Heat Transport and Afterheat Removal for GCRs under Accident Conditions (CRP-3)'. America, China, France, Germany, Japan, Netherlands and Russia participate the program. Final goal of the program is to show clearly to the world one of the most important salient features of the HTGR, that is the HTGR reactor can be cooled down by passive measures without causing any damage to the nuclear reactor system even in accidental conditions, and to make clear the boundaries (or restrictions) for the passive cooling regime. The first 5 year term of the coordinate program started in 1993 and established a goal to improve common knowledge for decay heat removal and to improve our tools, like computer codes and analytical models for the prediction of the performance of decay heat removal system. We are now performing benchmark problems for these purposes. The present efforts are concentrated on the benchmark for the passive heat removal performance outside the reactor vessel, partly because we have two different type of the HTGR in the world, the pebble bed type and the block type reactor. They have quite different heat dissipation behavior inside the reactor vessel. However, they have quite similar residual heat removal process outside the reactor vessel. For the first step of the international cooperation, we selected the common problem. After finishing the present benchmark we are planning to proceed to tackle the inside heat removal problem. (J.P.N.)

  6. Residual heat removal pump retrofit program

    International Nuclear Information System (INIS)

    Dudiak, J.G.; McKenna, J.M.

    1990-01-01

    Residual Heat Removal (RHR) pumps installed in pressurized water reactor power plants are used to provide the removal of decay heat from the reactor and to provide low head safety injection in the event of loss of coolant in the reactor coolant system. These pumps are subjected to rather severe temperature and pressure transients, therefore, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. RHR pumps have traditionally been a significant maintenance item for many utilities. The close-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. The casing separation requires the loosening of numerous highly torqued studs. Once the casing is separated, the impeller is dropped from the motor shaft to allow removal of the mechanical seal and casing cover from the motor shaft. Galling of the impeller to the motor shaft is not uncommon. The RHR pump internals are radioactive and the separation of the pump casing to perform routine maintenance exposes the maintenance personnel to high radiation levels. The handling of the impeller also exposes the maintenance personnel to high radiation levels. This paper introduces a design modification developed to convert the close-coupled RHR pumps to a coupled configuration

  7. Heat transfer and flow characteristics of a cooling thimble in a molten salt reactor residual heat removal system

    Directory of Open Access Journals (Sweden)

    Zonghao Yang

    2017-12-01

    Full Text Available In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

  8. An analysis of heat removal during cryogen spray cooling and effects of simultaneous airflow application.

    Science.gov (United States)

    Torres, J H; Tunnell, J W; Pikkula, B M; Anvari, B

    2001-01-01

    Cryogen spray cooling (CSC) is a method used to protect the epidermis from non-specific thermal injury that may occur as a result of various dermatological laser procedures. However, better understanding of cryogen deposition and skin thermal response to CSC is needed to optimize the technique. Temperature measurements and video imaging were carried out on an epoxy phantom as well as human skin during CSC with and without simultaneous application of airflow which was intended to accelerate cryogen evaporation from the substrate surface. An inverse thermal conduction model was used to estimate heat flux and total heat removed. Lifetime of the cryogen film deposited on the surface of skin and epoxy phantom lasted several hundred milliseconds beyond the spurt, but could be reduced to the spurt duration by application of airflow. Values over 100 J/cm(3) were estimated for volumetric heat removed from the epidermis using CSC. "Film cooling" instead of "evaporative cooling" appears to be the dominant mode of CSC on skin. Estimated values of heat removed from the epidermis suggest that a cryogen spurt as long as 200 milliseconds is required to counteract heat generated by high laser fluences (e.g., in treatment of port wine stains) in patients with high concentration of epidermal melanin. Additional cooling beyond spurt termination can be avoided by simultaneous application of airflow, although it is unclear at the moment if avoiding the additional cooling would be beneficial in the actual clinical situation. Copyright 2001 Wiley-Liss, Inc.

  9. Shutdown risk analysis for a BWR plant (residual heat removal systems)

    International Nuclear Information System (INIS)

    Rebollo Garcia, C.; Merino Teillet, A.; Cerezo, L.

    1994-01-01

    This report analyses the different risk situations which may arise during refuelling outage at Cofrentes NPP. The most critical situations are determined in terms of the small amount of coolant available and the lowest number of heat removal and water make-up systems available. The available times before the boiling point of the coolant is reached and the subsequent moment when the fuel elements are left uncovered in the event of the failure of the normal heat removal functions are determined. The analysis identifies the alternative systems which can be used besides those required by the technical specification and their capacity for residual heat removal and coolant make-up functions. (Author)

  10. Results from evaporation tests to support the MWTF heat removal system design

    International Nuclear Information System (INIS)

    Crea, B.A.

    1994-01-01

    An experimental tests program was conducted to measure the evaporative heat removal from the surface of a tank of simulated waste. The results contained in this report constitute definition design data for the latest heat removal function of the MWTF primary ventilation system

  11. Passive decay heat removal from the core region

    International Nuclear Information System (INIS)

    Hichen, E.F.; Jaegers, H.

    2002-01-01

    The decay heat in commercial Light Water Reactors is commonly removed by active and redundant safety systems supported by emergency power. For advanced power plant designs passive safety systems using a natural circulation mode are proposed: several designs are discussed. New experimental data gained with the NOKO and PANDA facilities as well as operational data from the Dodewaard Nuclear Power Plant are presented and compared with new calculations by different codes. In summary, the effectiveness of these passive decay heat removal systems have been demonstrated: original geometries and materials and for the NOKO facility and the Dodewaard Reactor typical thermal-hydraulic inlet and boundary conditions have been used. With several codes a good agreement between calculations and experimental data was achieved. (author)

  12. AEA studies on passive decay heat removal in advanced reactors

    International Nuclear Information System (INIS)

    Lillington, J.N.

    1994-01-01

    The main objectives of the UK study were: to identify, describe and compare different types of systems proposed in current designs; to identify key scenarios in which passive decay heat removal systems play an important preventative or mitigative role; to assess the adequacy of the relevant experimental database; to assess the applicability and suitability of current generation models/codes for predicting passive decay heat removal; to assess the potential effectiveness of different systems in respect of certain key licensing questions

  13. Heat removal in gas-cooled fuel rod clusters

    International Nuclear Information System (INIS)

    Rehme, K.

    1975-01-01

    For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

  14. Heat transfer augmentation for high heat flux removal in rib-roughened narrow channels

    International Nuclear Information System (INIS)

    Islam, M.S.; Hino, Ryutaro; Haga, Katsuhiro; Sudo, Yukio; Monde, Masanori.

    1997-03-01

    Heat transfer augmentation in narrow rectangular channels in a target system is a very important method to remove high heat flux up to 12 MW/m 2 generated at target plates of a high-intensity proton accelerator of 1.5 GeV and 1 mA with a proton beam power of 1.5 MW. In this report, heat transfer coefficients and friction factors in narrow rectangular channels with one-sided rib-roughened surface were evaluated for fully developed flows in the range of the Reynolds number from 6,000 to 1,00,000; the rib pitch-to-height ratios (p/k) were 10,20 and 30; the rib height-to-equivalent diameter ratios (k/De) were 0.025, 0.03 and 0.1 by means of previous existing experimental correlations. The rib-roughened surface augmented heat transfer coefficients approximately 4 times higher than the smooth surface at Re=10,000, p/k=10 and k/De=0.1; friction factors increase around 22 times higher. In this case, higher heat flux up to 12 MW/m 2 could be removed from the rib-roughened surface without flow boiling which induces flow instability; but pressure drop reaches about 1.8 MPa. Correlations obtained by air-flow experiments have showed lower heat transfer performance with the water-flow conditions. The experimental apparatus was proposed for further investigation on heat transfer augmentation in very narrow channels under water-flow conditions. This report presents the evaluation results and an outline of the test apparatus. (author)

  15. Post-accident fuel relocation and heat removal in the LMFBR

    International Nuclear Information System (INIS)

    Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

    1976-08-01

    Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

  16. Removal of contaminated concrete surfaces by microwave heating: Phase 1 results

    International Nuclear Information System (INIS)

    White, T.L.; Grubb, R.G.; Pugh, L.P.; Foster, D. Jr.; Box, W.D.

    1992-01-01

    Oak Ridge National Laboratory (ORNL) is developing a microwave heating process to remove radiologically contaminated surface layers from concrete. The microwave energy is directed at the concrete surface and heats the concrete and free water present in the concrete matrix. Continued heating produces steam-pressure-induced mechanical stresses that cause the concrete surface to burst. The concrete particles from this steam explosion are small enough to be removed by a vacuum system, yet less than 1% of the debris is small enough to pose an airborne contamination hazard. The first phase of this program has demonstrated reliable removal of noncontaminated concrete surfaces at frequencies of 2.45 GHz and 10.6 GHz. Continuous concrete removal rates of 1.07 cm 3 /s with 5.2 kW of 2.45.-GHz power and 2.11 cm 3 /s with 3.6 kW of 10.6-GHz power have been demonstrated. Figures-of-merit for microwave removal of concrete have been calculated to be 0.21 cm 3 /s/kW at 2.45 GHz and 0.59 cm 3 /s/kW at 10.6 GHz. The amount of concrete removed in a single pass can be controlled by choosing the frequency and power of the microwave system

  17. Numerical investigation of passive heat removal system via steam generator in VVER 1200

    International Nuclear Information System (INIS)

    Dinh Anh Tuan; Duong Thanh Tung; Tran Chi Thanh; Nguyen Van Thai

    2015-01-01

    Passive heat removal system (PHRS) via Steam Generator is an important part in VVER design. In case of Design Basic Accidents such as blackout, failure of feed water supply to steam generator or coolant leakage with failure of emergency core cooling at high pressure. PHRS is designed to remove the residual heat from reactor core through steam generator to heat exchanger which is placed outside reactor vessel. In order to evaluate the passive system, a numerical investigation using a CFD code is performed. However, PHRS has complex geometry for using CFD simulation. Thus, RELAP5 is applied to provide the wall heat flux of tube in the heat exchanger tank. The natural convection in the heat exchanger tank is investigated in this report. Numerical results show temperature and velocity distribution in the heat exchanger tank are calculated with different wall heat flux corresponding to various transient conditions. The calculated results contribute to the capacity analysis of passive heat removal system and giving valuable information for safe operation of VVER 1200. (author)

  18. Tests for removal of decay heat by natural convection

    International Nuclear Information System (INIS)

    Kashiwagi, E.; Wataru, M.; Gomi, Y.; Hattori, Y.; Ozaki, S.

    1993-01-01

    Interim storage technology for spent fuel by dry storage casks have been investigated. The casks are vertically placed in a storage building. The decay heat is removed from the outer cask surface by natural convection of air entering from the building wall to the roof. The air flow pattern in the storage building was governed by the natural driving pressure difference and circulating flow. The purpose of this study is to understand the mechanism of the removal of decay heat from casks by natural convection. The simulated flow conditions in the building were assumed as a natural and forced combined convection and were investigated by the turbulent quantities near wall. (author)

  19. Multiple pollutant removal using the condensing heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Jankura, B. J. [McDermott Technology Inc., Alliance, OH (United States); Kudlac, G. A. [McDermott Technology Inc., Alliance, OH (United States); Bailey, R. T. [McDermott Technology Inc., Alliance, OH (United States)

    1998-06-01

    The Integrated Flue Gas Treatment (IFGT) system is a new concept whereby a Teflon ® covered condensing heat exchanger is adapted to remove certain flue gas constituents, both particulate and gaseous, while recovering low level heat. The pollutant removal performance and durability of this device is the subject of a USDOE sponsored program to develop this technology. The program was conducted under contract to the United States Department of Energy's Fossil Energy Technology Center (DOE-FETC) and was supported by the Ohio Coal Development Office (OCDO) within the Ohio Department of Development, the Electric Power Research Institute's Environmental Control Technology Center (EPRI-ECTC) and Babcock and Wilcox - a McDermott Company (B&W). This report covers the results of the first phase of this program. This Phase I project has been a two year effort. Phase I includes two experimental tasks. One task dealt principally with the pollutant removal capabilities of the IFGT at a scale of about 1.2MWt. The other task studied the durability of the Teflon ® covering to withstand the rigors of abrasive wear by fly ash emitted as a result of coal combustion. The pollutant removal characteristics of the IFGT system were measured over a wide range of operating conditions. The coals tested included high, medium and low-sulfur coals. The flue gas pollutants studied included ammonia, hydrogen chloride, hydrogen fluoride, particulate, sulfur dioxide, gas phase and particle phase mercury and gas phase and particle phase trace elements. The particulate removal efficiency and size distribution was investigated. These test results demonstrated that the IFGT system is an effective device for both acid gas absorption and fine particulate collection. Although soda ash was shown to be the most effective reagent for acid gas absorption, comparative cost analyses suggested that magnesium enhanced lime was the most promising avenue for future study. The durability of the

  20. Analysis of a convection loop for GFR post-LOCA decay heat removal

    International Nuclear Information System (INIS)

    Williams, W.C.; Hejzlar, P.; Saha, P.

    2004-01-01

    A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA gas-cooled fast reactor (GFR). The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO 2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO 2 outdoes helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops. (authors)

  1. Antimony contamination, consequences and removal techniques: A review.

    Science.gov (United States)

    Li, Jiayu; Zheng, BoHong; He, Yangzhuo; Zhou, Yaoyu; Chen, Xiao; Ruan, Shan; Yang, Yuan; Dai, Chunhao; Tang, Lin

    2018-07-30

    A significant amount of antimony (Sb) enters into the environment every year because of the wide use of Sb compounds in industry and agriculture. The exposure to Sb, either direct consumption of Sb or indirectly, may be fatal to the human health because both antimony and antimonide are toxic. Firstly, the introduction of Sb chemistry, distribution and health threats are presented in this review, which is essential to the removal techniques. Then, we provide the recent and common techniques to remove Sb, including adsorption, coagulation/flocculation, membrane separation, electrochemical methods, ion exchange and extraction. Removal techniques concentrate on the advantages, drawbacks, economical efficiency and the recent achievements of each technique. We also take an overall consideration of experimental conditions, comparison criteria, and economic aspects. Copyright © 2018 Elsevier Inc. All rights reserved.

  2. Development of a new decay heat removal system for a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sim, Yoon Sub; Park, Rae Young; Kim, Seyun

    2007-01-01

    The heat removal capacity of a RCCS is one of the major parameters limiting the capacity of a HTGR based on a passive safety system. To improve the plant economy of a HTGR, the decay heat removal capacity needs to be improved. For this, a new analysis system of an algebraic method for the performance of various RCCS designs was set up and the heat transfer characteristics and performance of the designs were analyzed. Based on the analysis results, a new passive decay heat removal system with a substantially improved performance, LFDRS was developed. With the new system, one can have an expectation that the heat removal capacity of a HTGR could be doubled

  3. Cryogen spray cooling: Effects of droplet size and spray density on heat removal.

    Science.gov (United States)

    Pikkula, B M; Torres, J H; Tunnell, J W; Anvari, B

    2001-01-01

    Cryogen spray cooling (CSC) is an effective method to reduce or eliminate non-specific injury to the epidermis during laser treatment of various dermatological disorders. In previous CSC investigations, fuel injectors have been used to deliver the cryogen onto the skin surface. The objective of this study was to examine cryogen atomization and heat removal characteristics of various cryogen delivery devices. Various cryogen delivery device types including fuel injectors, atomizers, and a device currently used in clinical settings were investigated. Cryogen mass was measured at the delivery device output orifice. Cryogen droplet size profiling for various cryogen delivery devices was estimated by optically imaging the droplets in flight. Heat removal for various cryogen delivery devices was estimated over a range of spraying distances by temperature measurements in an skin phantom used in conjunction with an inverse heat conduction model. A substantial range of mass outputs were measured for the cryogen delivery devices while heat removal varied by less than a factor of two. Droplet profiling demonstrated differences in droplet size and spray density. Results of this study show that variation in heat removal by different cryogen delivery devices is modest despite the relatively large difference in cryogen mass output and droplet size. A non-linear relationship between heat removal by various devices and droplet size and spray density was observed. Copyright 2001 Wiley-Liss, Inc.

  4. Summary report of RAMONA investigations into passive decay heat removal

    International Nuclear Information System (INIS)

    Hoffmann, H.; Marten, K.; Weinberg, D.; Frey, H.H.; Rust, K.; Ieda, Y.; Kamide, H.; Ohshima, H.; Ohira, H.

    1995-07-01

    An important safety feature of an advanced sodium-cooled reactor (e.g. European Fast Reactor, EFR) is the passive decay heat removal. This passive concept is based on several direct reactor cooling systems operating independently from each other. Each of the systems consists of a sodium/sodium decay heat exchanger immersed in the primary vessel and connected via an intermediate sodium loop to a heat sink formed by a sodium/air heat exchanger installed in a stack with air inlet and outlet dampers. The decay heat is removed by natural convection on the sodium side and natural draft on the air side. To demonstrate the coolability of the pool-type primary system by buoyancy-driven natural circulation, tests were performed under steady-state and transient conditions in facilities of different scale and detail. All these investigations serve to understand the physical processes and to verify computer codes used to transfer the results to reactor conditions. RAMONA is the three-dimensional 1:20-scaled apparatus equipped with all active components. Water is used as simulant fluid for sodium. The maximum core power is 75 kW. The facility is equipped with about 250 thermocouples to register fluid temperatures. Velocities and mass flows are measured by Laser Doppler Anemometers and magneto-inductive flowmeters. Flow paths are visualized by tracers. The conclusion of the investigations is that the decay heat can be removed from the primary system by means of natural convection. Always flow paths develop, which ensure an effective cooling of all regions. This is even proved for extreme conditions, e.g. in case of delays of the decay heat exchanger startup, failures of several DHR chains, and a drop of the fluid level below the inlet windows of the IHXs and decay heat exchangers. (orig.) [de

  5. Numerical analysis of cavitating flow characteristics in impeller of residual heat removal pump

    NARCIS (Netherlands)

    Hong, Feng; Yuan, Jianping; Zhou, Banglun

    2016-01-01

    In order to investigate internal cavitating flow characteristics of the impeller in residual heat removal pumps, the three-dimensional cavitating flow in a residual heat removal model pump is numerically calculated by using the homogeneous mixture cavitation model based on the Rayleigh-Plesset

  6. Mathematical modelling for magnetite (crude removal from primary heat transfer loop by ion-exchange resins

    Directory of Open Access Journals (Sweden)

    Zeeshan Nawaz

    2009-04-01

    Full Text Available The present research focuses to develop mathematical model for the removal of iron (magnetite by ion-exchange resin from primary heat transfer loop of process industries. This mathematical model is based on operating capacities (that’s provide more effective design as compared to loading capacity from static laboratory tests. Results showed non-steady state distribution of external Fe2+ and limitations imposed on operating conditions, these conditions includes; loading and elution cycle time, flow rate, concentration of both loading and removal, volume of resin required. Number of generalized assumptions was made under shortcut modeling techniques to overcome the gap of theoretical and actual process design.

  7. Removal of decay heat by specially designed isolation condensers for advanced heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dhawan, M L; Bhatia, S K [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    For Advanced Heavy Water Reactor (AHWR), removal of decay heat and containment heat is being considered by passive means. For this, special type of isolation condensers are designed. Isolation condensers when submerged in a pool of water, are the best choice because condensation of high temperature steam is an extremely efficient heat transfer mechanism. By the use of isolation condensers, not only heat is removed but also pressure and temperature of the system are automatically controlled without losing the coolant and without using conventional safety relief valves. In this paper, design optimisation studies of isolation condensers of different types with natural circulation for the removal of core decay heat for AHWR is presented. (author). 8 refs., 2 figs.

  8. Impact of the amount of working fluid in loop heat pipe to remove waste heat from electronic component

    Directory of Open Access Journals (Sweden)

    Smitka Martin

    2014-03-01

    Full Text Available One of the options on how to remove waste heat from electronic components is using loop heat pipe. The loop heat pipe (LHP is a two-phase device with high effective thermal conductivity that utilizes change phase to transport heat. It was invented in Russia in the early 1980’s. The main parts of LHP are an evaporator, a condenser, a compensation chamber and a vapor and liquid lines. Only the evaporator and part of the compensation chamber are equipped with a wick structure. Inside loop heat pipe is working fluid. As a working fluid can be used distilled water, acetone, ammonia, methanol etc. Amount of filling is important for the operation and performance of LHP. This work deals with the design of loop heat pipe and impact of filling ratio of working fluid to remove waste heat from insulated gate bipolar transistor (IGBT.

  9. Study on constraints for heat removal duties of the main fractionator in delayed coking units

    International Nuclear Information System (INIS)

    Lei, Yang; Zhang, Bingjian; Qi, Xin; Chen, Qinglin; Hui, Chi-Wai

    2014-01-01

    A novel method is presented in this paper to quantitatively define the heat removal of the main fractionator in delayed coking units on the basis of a fractionating precision diagram (Houghland diagram) and column grand composite curve (CGCC). By referring to the CGCC method, several envelopes are illustrated at draw trays including the top pumparound draw, diesel draw, intermediate pumparound draw and gas oil draw, the energy and material balances are then calculated. Assuming practical near-minimum thermodynamic condition (PNMTC), the minimum liquid reflux flow is zero in the envelope for pumparound trays without product draw and the minimum liquid reflux flow is defined by Houghland diagram for pumparound trays with product draw. The PNMTC-CGCC is constructed by calculating the enthalpy-flow deficit to quantitatively define the heat removal constraints in each envelope. Meanwhile, the corresponding practical heat removal curve is constructed. A case study shows that the high temperature heat removal ratio within the main fractionator increased by 8%. The proposed method offers heat removal inequality constraints for the model to optimize the heat integration between the main fractionator and the heat exchanger network. - Highlights: • A novel method defines the heat removal constraints of the main fractionator. • Fractionating precision diagram and column grand composite curve are combined. • The results are the inequality constraints in a simultaneous optimization model

  10. New atraumatic easy removal technique for permanently cemented crown

    Directory of Open Access Journals (Sweden)

    Pravinkumar G Patil

    2012-01-01

    Full Text Available Removal of a permanently cemented crown or fixed partial denture is a cumbersome procedure for a prosthodontist, especially when there is no purchase point available to remove it. The technique described in this article consists of sectioning of a crown on facial surface followed by removal of the crown with orthodontic plier. This technique does not damage the gingival/periodontal tissues or underlying tooth structure as the crown need not to be removed with jerky back-action force.

  11. Investigation of characteristics of passive heat removal system based on the assembled heat transfer tube

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Xiang Cheng; Yan, Changqi; Meng, Zhao Ming; Chen, Kailun; Song, Shao Chuang; Yang, Zong Hao; Yu, Jie [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2016-12-15

    To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from 450 .deg. C to 700 .deg. C and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

  12. Investigation of Characteristics of Passive Heat Removal System Based on the Assembled Heat Transfer Tube

    Directory of Open Access Journals (Sweden)

    Xiangcheng Wu

    2016-12-01

    Full Text Available To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from 450°C to 700°C and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

  13. Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

    International Nuclear Information System (INIS)

    WILLIS, W.L.

    2000-01-01

    This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein

  14. Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

    Energy Technology Data Exchange (ETDEWEB)

    WILLIS, W.L.

    2000-06-15

    This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein.

  15. Analysis of decay heat removal by natural convection in PFBR

    International Nuclear Information System (INIS)

    Kasinathan, N.; Vaidyanathan, G.; Chetal, S.C.; Bhoje, S.B.

    1993-01-01

    PFBR is a 500 MWe, 1200 MWt pool type LMFBR. In order to assure reliable decay heat removal, four totally independent Safety Grade Decay Heat Removal Systems (SGDHRS) which removes heat directly from the hot pool, is provided. Each of the SGDHRS comprises of a hot pool dipped decay heat exchanger (DHX), a sodium - air heat exchanger (AHX) at a suitable elevation and associated piping and circuits. This paper brings out the step by step approach that have been taken to decide on the preliminary sizing of the SGDHRS components, and static and transient analysis to assess the adequacy of the Decay Heat Removal capacity of the SGDHRS during the worst of the foreseen design basis conditions. The maximum values the important safety related temperatures viz., clad hotspot, hot pool top surface, reactor inlet, fuel subassembly outlets etc., would reach, can be obtained only through a comprehensive transient analysis. In order to get quick and reasonably meaningful results, one dimensional thermal-hydraulics models for the core, hot and cold pools, IHX, DHX, AHX and various pipings were developed and a code DHDYN formulated. With this a total power failure situation followed by initiations of DHR half an hour later was studied and the results revealed the following: (i) clad hotspot temperature in the in-vessel stored spent fuel subassemblies could be held below 800 deg. C only if primary sodium flow through these subassemblies are increased up to three times the originally allocated flow in the design, (ii) hotpool top zone temperature reaches 572 deg. C, (iii) reactor inlet temperature reaches 482 deg. C, (iv) the hot pool top zone temperature cools down to 450 deg. C in about 25 h. Thus these results satisfactorily established the adequacy of the sizing and the capability of the SGDHRS. DHDYN code is also used to study the RAMONA water experiments conducted in Germany. Initial results available has brought out the conservative nature of the DHDYN predictions as compared

  16. Heat removing device for reactor container

    International Nuclear Information System (INIS)

    Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

    1996-01-01

    A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

  17. Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

    2015-06-15

    The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

  18. Passive deca-heat removal in the fixed bed nuclear reactor (FBNR) - 15551

    International Nuclear Information System (INIS)

    Solano Diaz, E.C.; Luna Aguilera, G.M.; Santos, R.A.; Vaca, D.E.

    2015-01-01

    The Fixed Bed Nuclear Reactor (FBNR) is a Generation IV small reactor concept, where the spherical elements contain Triso-type microspheres with UO 2 , which serves as nuclear fuel. In the event that adverse operation conditions occur, the water pump is automatically shut off and the fuel pebbles fall back by gravity into the fuel chamber. Since the FBNR relies on passive security systems, the removal of the decay heat in the fuel chamber is achieved by contact with quiescent water. In the present paper, a mathematical simulation of the passive cooling of the system was conducted in SOLIDWORKS so as to obtain a temperature profile in the body during the decay heat removal process. Homogenization techniques were employed to smooth out spatial variations across the multiphase system and to derive expression for the effective thermophysical properties that are valid through the macroscopic entry (the chamber). The simulation showed that the chamber's temperature rose from 573 K to its maximum temperature, 1234 K, in the first hour. Afterwards, the temperature fluctuated, but stayed under 552 K. Since the temperature of the system was always kept under the value of the safety parameter (1200 C. degrees) the simulation confirmed that an effective passive cooling of the fuel chamber is indeed feasible. (authors)

  19. Experimental and analytical studies of a passive shutdown heat removal system for advanced LMRs

    International Nuclear Information System (INIS)

    Heineman, J.; Kraimer, M.; Lottes, P.; Pedersen, D.; Stewart, R.; Tessier, J.

    1988-01-01

    A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance

  20. Experimental and analytical studies of a passive shutdown heat removal system for advanced LMRs

    Energy Technology Data Exchange (ETDEWEB)

    Heineman, J.; Kraimer, M.; Lottes, P.; Pedersen, D.; Stewart, R.; Tessier, J.

    1988-01-01

    A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance.

  1. Feasibility of passive heat removal systems

    Energy Technology Data Exchange (ETDEWEB)

    Ashurko, Yu M [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1996-12-01

    This paper presents a review of decay heat removal systems (DHRSs) used in liquid metal-cooled fast reactors (LMFRs). Advantages and the disadvantages of these DHRSs, extent of their passivity and prospects for their use in advanced fast reactor projects are analyzed. Methods of extending the limitations on the employment of individual systems, allowing enhancement in their effectiveness as safety systems and assuring their total passivity are described. (author). 10 refs, 10 figs.

  2. Specialists' meeting on evaluation of decay heat removal by natural convection

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-02-01

    Decay heat removal by natural convection (DHRNC) is essential to enhancing the safety of liquid metal fast reactors (LMFRs). Various design concepts related to DHRNC have been proposed and experimental and analytical studies have been carried out in a number of countries. The purpose of this Specialists' Meeting on 'Decay Heat Removal by Natural Convection' organized by the International Working Group on Fast Reactors IAEA, is to exchange information about the state of the art related to methodologies on evaluation of DHRNC features (experimental studies and code developments) and to discuss problems which need to be solved in order to evaluate DHRNC properly and reasonably. The following main topical areas were discussed by delegates: Overview; Experimental studies and code validation; Design study. Two main DHR systems for LMFR are under consideration: (i) direct reactor auxiliary cooling system (DRACS) with immersed DFIX in main vessel, intermediate sodium loop and sodium-air heat exchanger; and (ii) auxiliary cooling system which removes heat from the outside surface of the reactor vessel by natural convection of air (RVACS). The practicality and economic viability of the use of RVACS is possible up to a modular type reactor or a middle size reactor based on current technology. For the large monolithic plant concepts DRACS is preferable. The existing experimental results and the codes show encouraging results so that the decay heat removal by pure natural convection is feasible. Concerning the objective, 'passive safety', the DHR by pure natural convection is essential feature to enhance the reliability of DHR.

  3. Specialists' meeting on evaluation of decay heat removal by natural convection

    International Nuclear Information System (INIS)

    1993-02-01

    Decay heat removal by natural convection (DHRNC) is essential to enhancing the safety of liquid metal fast reactors (LMFRs). Various design concepts related to DHRNC have been proposed and experimental and analytical studies have been carried out in a number of countries. The purpose of this Specialists' Meeting on 'Decay Heat Removal by Natural Convection' organized by the International Working Group on Fast Reactors IAEA, is to exchange information about the state of the art related to methodologies on evaluation of DHRNC features (experimental studies and code developments) and to discuss problems which need to be solved in order to evaluate DHRNC properly and reasonably. The following main topical areas were discussed by delegates: Overview; Experimental studies and code validation; Design study. Two main DHR systems for LMFR are under consideration: (i) direct reactor auxiliary cooling system (DRACS) with immersed DFIX in main vessel, intermediate sodium loop and sodium-air heat exchanger; and (ii) auxiliary cooling system which removes heat from the outside surface of the reactor vessel by natural convection of air (RVACS). The practicality and economic viability of the use of RVACS is possible up to a modular type reactor or a middle size reactor based on current technology. For the large monolithic plant concepts DRACS is preferable. The existing experimental results and the codes show encouraging results so that the decay heat removal by pure natural convection is feasible. Concerning the objective, 'passive safety', the DHR by pure natural convection is essential feature to enhance the reliability of DHR

  4. Device for removing alkali metal residues from heat exchanger

    International Nuclear Information System (INIS)

    Matal, O.

    1987-01-01

    The main parts of the facility consists of a condensing vessel and a vacuum pump unit interconnected via a vacuum pipe. The heat exchanger is heated to a temperature at which the alkali metal residues evaporate. Metal vapors are collected in the condensing vessel where they condense. The removal of the alkali metal residues from the heat exchanger pipes allows thorough inspection of the pipe inside during scheduled nuclear power plant shutdowns. The facility can be used especially with reverse steam generators. (E.S.). 1 fig

  5. After heat removing system of a nuclear reactor

    International Nuclear Information System (INIS)

    Hayashi, Takao; Yamada, Masao; Ohashi, Kazutaka.

    1994-01-01

    In a variable conductance heat pipe of an after heat removing system, an evaporation portion and a condensator are connected by a steam diffusing path for an operation fluid and a liquid condensate recycling path. Further, incondensible gases are sealed at the inside together with the operation fluid, and a gas reservoir for the incondensible gases is disposed at the downstream of a condensation portion. If heat input is applied to the evaporation portion of the heat pipe, the incondensible gases are separated to form a boundary between both of them. When the amount of heat applied is small, the incondensible gases partially seal the condensation portion to form a local condensation insensitive portion, so that a heat conductance can be suppressed low. On the other hand, as the amount of heat inputted is increased, the incondensible gases are compressed, the heat conduction area of the condensation portion is increased and a heat conductance is increased to conduct self-control so as to increase heat transfer performance of the heat pipe. Then, the liquid condensate is recycled to the evaporation portion by spontaneous dripping of the condensate itself without wick, thereby enabling to conduct automatic switching so as to increase the heat dissipation amount to maximum. (N.H.)

  6. Evaporation and condensation devices for passive heat removal systems in nuclear power engineering

    International Nuclear Information System (INIS)

    Gershuni, A.N.; Pis'mennyj, E.N.; Nishchik, A.P.

    2016-01-01

    The paper justifies advantages of evaporation and condensation heat transfer devices as means of passive heat removal and thermal shielding in nuclear power engineering. The main thermophysical factors that limit heat transfer capacity of evaporation and condensation systems have been examined in the research. The results of experimental studies of heat engineering properties of elongated (8-m) vertically oriented evaporation and condensation devices (two-phase thermosyphons), which showed a high enough heat transfer capacity, as well as stability and reliability both in steady state and in start-up modes, are provided. The paper presents the examples of schematic designs of evaporation and condensation systems for passive heat removal and thermal shielding in application to nuclear power equipment

  7. Excessive heat removal due to feedwater system malfunction

    International Nuclear Information System (INIS)

    Beader, D.; Peterlin, G.

    1986-01-01

    Excessive heat removal transient of the Krsko Nuclear Power Plant, caused by steam generators feedwater system malfunctions was simulated by RELAP5/MOD1 computer code. The results are increase of power and reactor scram caused by high-high steam generator level. (author)

  8. Hydrodynamical tests with an original PWR heat removal pump

    International Nuclear Information System (INIS)

    Wietstock, P.

    1984-01-01

    GKSS-Forschungszentrum performes hydrodynamical tests with an original PWR heat removal pump to analyse the influences of fluid parameters on the capacity and cavitation behavior of the pump in order to get further improvements of the quantification of the reached safety-level. It can be concluded, that in case of the tested heat removal pump the additional loads during transition from cavitation free operation into fully cavitation for the investigated operation point with 980 m 3 /h will be smaller than the alteration of loads during passing through the total characteristic. The results from cavitation tests for other operation points indicate, that this very important consequence especially for accident operation will be valid for the total specified pump flow area. (orig.)

  9. String Technique for Anterior Orbital Fish Hook Removal.

    Science.gov (United States)

    Starr, Matthew R; Choi, Michael B; Mahr, Michael A; Mettu, Pradeep; Patterson, David F

    2018-06-13

    Removing fish hooks is a common procedure performed by many emergency department providers. There are several techniques that are commonly employed to aid in successful removal. However, when a fish hook becomes embedded within the orbit, there are limited options as to avoid damaging vital surrounding structures. The authors report the removal of a fish hook within the anterior orbit using the string technique in a 25-year-old patient. The procedure was performed under general anesthesia with the aid of size 5 polyglactin suture wrapped around the hook. The procedure itself took less than 10 seconds and was successful in swiftly and safely removing the hook without damaging surrounding orbital structures. The patient recovered well without any permanent sequelae.

  10. Decay heat removal and transient analysis in accidental conditions in the EFIT reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Meloni, P.; Polidori, M.; Casamirra, M.; Castiglia, F.; Giardina, M.

    2007-01-01

    The development of a conceptual design of an industrial scale transmutation facility (EFIT) of several 100 MW thermal power based on Accelerator Driven System (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which lead to the Loss of Heat Sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1-D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios. (author)

  11. Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

    Directory of Open Access Journals (Sweden)

    Giacomino Bandini

    2008-01-01

    Full Text Available The development of a conceptual design of an industrial-scale transmutation facility (EFIT of several 100 MW thermal power based on accelerator-driven system (ADS is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS. In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

  12. Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor. Comparisons of the decay heat removal characteristics on lead, lead-bismuth and sodium cooled reactors

    International Nuclear Information System (INIS)

    Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

    2000-04-01

    The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. In this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube failure accidents in a steam generator. In this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in Equivalent plant' with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. In conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to confirm the heat transfer reduction by the oxidized film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance. (author)

  13. Confirmatory analysis of the AP1000 passive residual heat removal heat exchanger with 3-D computational fluid dynamic analysis

    International Nuclear Information System (INIS)

    Schwall, James R.; Karim, Naeem U.; Thakkar, Jivan G.; Taylor, Creed; Schulz, Terry; Wright, Richard F.

    2006-01-01

    The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power up-rate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model

  14. Heating Changes Bio-Schwertmannite Microstructure and Arsenic(III Removal Efficiency

    Directory of Open Access Journals (Sweden)

    Xingxing Qiao

    2017-01-01

    Full Text Available Schwertmannite (Sch is an efficient adsorbent for arsenic(III removal from arsenic(III-contaminated groundwater. In this study, bio-schertmannite was synthesized in the presence of dissolved ferrous ions and Acidithiobacillus ferrooxidans LX5 in a culture media. Bio-synthesized Sch characteristics, such as total organic carbon (TOC, morphology, chemical functional groups, mineral phase, specific surface area, and pore volume were systematically studied after it was dried at 105 °C and then heated at 250–550 °C. Differences in arsenic(III removal efficiency between 105 °C dried-sch and 250–550 °C heated-sch also were investigated. The results showed that total organic carbon content in Sch and Sch weight gradually decreased when temperature increased from 105 °C to 350 °C. Sch partly transformed to another nanocrystalline or amorphous phase above 350 °C. The specific surface area of 250 °C heated-sch was 110.06 m2/g compared to 5.14 m2/g for the 105 °C dried-sch. Total pore volume of 105 °C dried-sch was 0.025 cm3/g with 32.0% mesopore and 68.0% macropore. However, total pore volume of 250 °C heated-mineral was 0.106 cm3/g with 23.6% micropore, 33.0% mesopore, and 43.4% macropore. The arsenic(III removal efficiency from an initial 1 mg/L arsenic(III solution (pH 7.5 was 25.1% when 0.25 g/L of 105 °C dried-sch was used as adsorbent. However, this efficiency increased to 93.0% when using 250 °C heated-sch as adsorbent. Finally, the highest efficiency for arsenic(III removal was obtained with sch-250 °C due to high amounts of sorption sites in agreement with the high specific surface area (SSA obtained for this sample.

  15. Study on concrete cask for practical use. Heat removal test under normal condition

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Wataru, Masumi; Shirai, Koji; Saegusa, Toshiari

    2005-01-01

    In Japan, it is planed to construct interim storage facilities taking account of dry storage away form reactor in 2010. Recently, a concrete cask is noticed from the economical point of view. But data for its safety analysis have not been sufficient yet. Heat removal tests using to types of full-scale concrete casks were conducted. This paper describes the results under normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced Concrete cask (RC cask) and Concrete Filled Steel cask (CFS cask) were the specimen casks. The levels of decay heat at the initial period of 60 years of storage, the intermediate period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data required for heat removal analyses were obtained. (author)

  16. Post-accident heat removal research: A state of the art review

    International Nuclear Information System (INIS)

    Mueller, U.; Schulenberg, T.

    1983-11-01

    For a realistic assessment of the consequence of extremely unlikely reactor accidents resulting in core degradation or core meltdown key questions are how to remove the decay heat from the reactor system and how to retain the radioactive core debris within the containment. Usually, this complex of questions is referred to as Post-Accident Heat Removal (PAHR). In this article the research work on PAHR performed by various institutions during the last decade has been reviewed. The main results have been summarized under the chapter headings ''Accident Scenarios,'' - ''Core Debris Accommodation Concepts,'' and ''PAHR Topics.'' Particular emphasis has been placed on the presentation of the following problems: characteristics and coolability of solid core debris in the vector vessel, heat removal from molten pools of core material, and core-melt interaction with structural materials. Some unresolved or insufficiently answered questions relating to special ''PAHR Topics'' have been mentioned or discussed at the end of the particular Chapter. Problem areas of major uncertainty have been identified and listed at the end of the review article. They include the following subjects: formation of debris beds and bed characteristics, post dryout behaviour of particle beds, long-term availability and proper location of heat sinks, creep rupture of structures under high thermal loads. (orig.) [de

  17. Passive decay heat removal by natural circulation

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Venkat Raj, V.; Kakodkar, A.; Mehta, S.K.

    1990-01-01

    The standardised 235 MWe PHWRs being built in India are the pressure tube type, heavy water moderated, heavy water cooled and natural uranium fuelled reactors. Several passive safety features are incorporated in these reactors. These include: (1) Containment pressure reduction and fission product trapping with the help of suppression pool following LOCA. (2) Emergency coolant injection by means of accumulators. (3) Large heat sink provided by the low temperature moderator under accident conditions. (4) Low excess reactivity, through the use of natural uranium fuel and on power fuelling. (5) Residual heat removal by means of natural circulation, etc. of which the last item is the subject matter of this report. (author). 8 refs, 10 figs

  18. Control of the ASTRA decay heat removal system

    International Nuclear Information System (INIS)

    Nedelik, A.

    1982-11-01

    To ensure a minimum of core cooling even under severest accident conditions (loss of reactor pool water) a core spray system for decay heat removal has been installed at the ASTRA-reactor. The automatic and manual control of the system, its power supply and test procedures are shortly described. (Author)

  19. Design of CAREM-25 Residual Heat Removal System: Nuclear Safety Aspects

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo; Schlamp, Miguel; Barrera, M.

    2000-01-01

    In this paper Carem-25 residual heat removal system (RHRS) design is analyzed from the nuclear safety point of view.The proposed RHRS is a condenser that transfers the heat to a pool located in the upper level of the containment.The RHRS design basis accident is a reactor loss of heat sink.The following requirements were settled to be verified: a) To remove 2 MW, for a primary circuit pressure of 12.25 MPa and a pool temperature of 100 0 C. b) No condenser tubes flooding, for a primary circuit pressure of 14 MPa and a pool temperature of 100 0 C. c) To reach hot shutdown in 48-hrs, that is to remove of 0.6 MW for a primary circuit pressure of 2.3 MPa and a pool temperature of 120 0 C.Heat transfer regimes inside and outside the condenser and flow patterns were analyzed.Steady state conditions for the above design conditions were modeled.The design requirements were verified taking into account heat transfer coefficients uncertainties and their propagation to the equipment elevation in the containment over the RPV, in order to minimize its elevation and its possible flooding.The resulting condenser tubes were 2 S CH 160 TP 347 SS, with a total area of 4 m 2 and a required minimum height of 6 m from the RPV water level to the condenser outlet headers

  20. Residual Heat Removal System qualitative probabilistic safety analysis before and after auto closure interlock removal

    International Nuclear Information System (INIS)

    Mikulicic, V.; Simic, Z.

    1992-01-01

    The analysis evaluates the consequences of the removal of the auto closure interlock (ACI) on the Residual Heat Removal System (RHRS) suction/isolation valves at the nuclear power plant. The deletion of the RHRS ACI is in part based on a probabilistic safety analysis (PSA) which justifies the removal based on a criterion of increased availability and reliability. Three different areas to be examined in PSA: the likelihood of an interfacing system LOCA; RHRS availability and reliability; and low temperature overpressurization control. The paper emphasizes particularly the RHRS unavailability and reliability evaluation utilizing the current control circuitry configuration and then with the proposed modification to the control circuitry. (author)

  1. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Goto, Tadashi; Inoue, Kotaro; Yamakawa, Masanori; Ikeda, Takashi.

    1988-01-01

    Purpose: To promote more positive forcive circulation of primary circuit fluids thereby increase the heat removing amount. Constitution: The primary side of an electromagnetic flow coupler type heat exchanger is opened to the primary fluid of a reactor, while the secondary side is connected with the secondary circuit comprising an air cooler and an electromagnetic pump. Since the secondary circuit stands-by during normal operation, the electromagnetic flow coupler does not operate and does not generate force for flowing primary circuit fluid. If flow due to the external force to the primary circuit fluid should occur in the electromagnetic flow coupler type heat exchanger, an electromagnetic force tending to flow the secondary circuit fluid is exerted oppositely. However the coupler undergoes reaction inertia of the fluid or flowing resistance, to exert in the direction of suppressing the flow, thereby prevent the heat loss. (Yoshihara, H.)

  2. Decay heat removal plan of the SNR-300: a licensed concept

    International Nuclear Information System (INIS)

    Morgenstern, F.H.; Gyr, W.; Stoetzel, H.; Vossebrecker, H.

    1976-01-01

    The report describes how the decay heat removal plan of the SNR-300 has been established in 3 essential licensing steps, thus giving a very significant example for the slow but steady progress in the overall licensing process of the plant. (1) Introduction of an ECCS in addition to the 3 main heat transfer chains as a back-up for rather unlikely and undefined occurrences, 1970; (2) Experimental and computational demonstration of a reliable functioning of the in-vessel natural convection of the fluid flow, 1974; and (3) Proof of fulfilling the general safety and specific reliability criteria for the overall decay heat removal plan; i.e., the 3 main heat transfer chains with specific installations on the steam/water system side and the ECCS, 1976. Some special problem areas, for instance the cavity concept provided for the pipe fracture accident, have still to be licensed, but they do not contribute considerably to the overall risk

  3. Development of evaluation method for heat removal design of dry storage facilities. pt. 1. Heat removal test on vault storage system of cross flow type

    International Nuclear Information System (INIS)

    Sakamoto, Kazuaki; Koga, Tomonari; Wataru, Masumi; Hattori, Yasuo

    1997-01-01

    The report describes the result of heat removal test of passive cooling vault storage system of cross flow type using 1/5 scale model. Based on a prospect of steady increase in the amount of spent fuel, it is needed to establish large capacity dry storage technologies for spent fuel. Air flow patterns, distributions of air temperature and velocity were measured, by which heat removal characteristics of the system were made clear. Air flow patterns in the storage module depended on the ratio of the buoyant force to the inertial force; the former generated by the difference of air temperatures and the height of the storage module, the latter by the difference of air densities between the outlet of the storage module and ambience and the height of the chimney of the storage facility. A simple method to estimate air flow patterns in the storage module was suggested, where Ri(Richardson) number was applied to represent the ratio. Moreover, heat transfer coefficient from a model of storage tube to cooling air was evaluated, and it was concluded that the generalized expression of heat transfer coefficient for common heat exchangers could be applied to the vault storage system of cross flow type, in which dozens of storage tubes were placed in a storage module. (author)

  4. Separately removable tubes in heavy duty heat exchanger assemblies

    International Nuclear Information System (INIS)

    Neudeck, G.T.

    1980-01-01

    The invention is directed to removable heat exchanger tube assemblies in heavy duty equipment radiators in which the tubes are each separately removable if they become defective in service. An inwardly facing annular ledge or abutment is molded into the inside diameter of each upper and lower sealing member to receive the respective ends of the tubes and prevent vertical movement of the tubes in service. A flange or shoulder is also provided on the lower portions of each tube and engages the inside of the lower sealing member to further restrain downward movement of the tubes in service. Each tube may be removed by pushing the tube upwardly to overcome the upper ledge abutment and thereby lift the tube free of the lower seal. Each tube may then be removed sidewise from the radiator. Variations of the removable sealing arrangement can be made and are described herein

  5. The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Xiong, Zhenqin [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Gu, Hanyang, E-mail: guhanyang@stu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Wang, Minglu [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Cheng, Ye [Shanghai Nuclear Engineering Research and Design Institute, Shanghai 200233 (China)

    2014-12-15

    Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m{sup 2}/s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10{sup −2} m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m{sup 2}/s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow

  6. The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

    International Nuclear Information System (INIS)

    Xiong, Zhenqin; Gu, Hanyang; Wang, Minglu; Cheng, Ye

    2014-01-01

    Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m 2 /s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10 −2 m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m 2 /s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow rate and

  7. A study on the characteristics of the decay heat removal capacity for a large thermal rated LMR design

    International Nuclear Information System (INIS)

    Uh, J. H.; Kim, E. K.; Kim, S. O.

    2003-01-01

    The design characteristics and the decay heat removal capacity according to the type of DHR (Decay Heat Removal) system in LMR are quantitatively analyzed, and the general relationship between the rated core thermal power and decay heat removal capacity is created in this study. Based on these analyses results, a feasibility of designing a larger thermal rating KALIMER plant is investigated in view of decay heat removal capacity, and DRC (Direct Reactor Cooling) type DHR system which rejects heat from the reactor pool to air is proper to satisfy the decay heat removal capacity for a large thermal rating plant above 1,000 MWth. Some defects, however, including the heat loss under normal plant operation and the lack of reliance associated with system operation should be resolved in order to adopt the total passive concept. Therefore, the new concept of DHR system for a larger thermal rating KALIMER design, named as PDRC (passive decay heat removal circuit), is established in this study. In the newly established concept of PDRC, the Na-Na heat exchanger is located above the sodium cold pool and is prevented from the direct sodium contact during normal operation. This total passive feature has the superiority in the aspect of the minimizing the normal heat loss and the increasing the operation reliance of DHR system by removing either any operator action or any external operation signal associated with system operation. From this study, it is confirmed that the new concept of PDRC is useful to the designing of a large thermal rating power plant of KALIMER-600 in view of decay heat removal capability

  8. Strategy of experimental studies in PNC on natural convection decay heat removal

    International Nuclear Information System (INIS)

    Ieda, Y.; Kamide, H.; Ohshima, H.; Sugawara, S.; Ninokata, H.

    1993-01-01

    Experimental studies have been and are being carried out in PNC to establish the design and safety evaluation methods and the design and safety evaluation guide lines for decay heat removal by natural convection. A strategy of the experimental studies in PNC is described in this paper. The sphere of studies in PNC is to develop the evaluation methods to be available to DRACS as well as PRACS and IRACS for the plant where decay heat is removed by natural convection in some cases of loss of station service power. Similarity parameters related to natural convection are derived from the governing equations. The roles of both sodium and water experiments are defined in consideration of the importance of the similarity parameters and characteristics of scale model experiments. The experimental studies in PNC are reviewed. On the basis of the experimental results, recommended evaluation methods are shown for decay heat removal feature by natural convection. Future experimental works are also proposed. (author)

  9. Design of Passive Decay Heat Removal System using Mercury Thermosyphon for SFR

    Energy Technology Data Exchange (ETDEWEB)

    You, Byung Hyun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, thermosyphon application is suggested to accomplish the fully passive safety grade system and compactness of components via enhance the heat removal performance. A two-phase evaporating thermosyphon operates when the evaporator is heated, the working fluid start boiling, the vapor that is formed moves to the condenser, where it is condensed on the walls, giving up the heat of phase change to the cooling fluid. Gravity forces cause the condensate to condensed liquid flow to the evaporator again. These processes occur continuously, which causes transfer of heat from evaporator to condenser vice versa. After the thermal design and performance evaluation, the results were compared with the performance of conventional DRACS system. For the same amount of decay heat removal performance of PDRC system of KALIMER-600 mercury thermosyphon system can archive around 30∼50% of compactness. For the detailed design, improved analytical model and experimental data for the validation will be required to specify the new DHR system.

  10. Method for removal of decay heat of radioactive substances

    International Nuclear Information System (INIS)

    Hesky, H.; Wunderer, A.

    1981-01-01

    In this process, the decay heat from radioactive substances is removed by means of a liquid carried in the coolant loop. The liquid is partially evaporated by the decay heat. The steam is used to drive the liquid through the loop. When a static pressure level equivalent to the pressure drop in the loop is exceeded, the steam is separated from the liquid, condensed, and the condensate is reunited with the return flow of liquid for partial evaporation. (orig.) [de

  11. Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

    Science.gov (United States)

    Bersano, Andrea

    With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be

  12. Transient testing of the FFTF for decay-heat removal by natural convection

    International Nuclear Information System (INIS)

    Beaver, T.R.; Johnson, H.G.; Stover, R.L.

    1982-06-01

    This paper reports on the series of transient tests performed in the FFTF as a major part of the pre-operations testing program. The structure of the transient test program was designed to verify the capability of the FFTF to safely remove decay heat by natural convection. The series culminated in a scram from full power to complete natural convection in the plant, simulating a loss of all electrical power. Test results and acceptance criteria related to the verification of safe decay heat removal are presented

  13. Nuclear reactor with makeup water assist from residual heat removal system

    International Nuclear Information System (INIS)

    Schulz, T.L.; Corletti, M.M.

    1994-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit by pumping water from an in-containment refueling water storage tank during staged depressurization of the coolant circuit, the final stage including passive emergency cooling by gravity feed from the refueling water storage tank to the coolant circuit and to flood the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and avoids the final stage of depressurization with its flooding of the containment when such action is not necessary, but does not prevent the final stage when it is necessary. A high pressure makeup water storage tank coupled to the reactor coolant circuit holds makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal system can also be coupled in a loop with the refueling water supply tanks for cooling the tank. (Author)

  14. Possibility of a pressurized water reactor concept with highly inherent heat removal following capability

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Murao, Yoshio

    1995-01-01

    If the core power inherently follows change in heat removal rate from the primary coolant system within small thermal expansion of the coolant which can be absorbed in a practical size of pressurizer, reactor systems may have more safety and load following capability. In order to know possibility and necessary conditions of a concept on reactor core and primary coolant system of a pressurized water reactor (PWR) with such 'highly inherent heat removal following capability', transient analyses on an ordinary two-loop PWR have been performed for a transient due to 50% change in heat removal with the RETRAN-02 code. The possibility of a PWR concept with the highly inherent heat removal following capability has been demonstrated under the conditions of the absolute value of ratio of the coolant density reactivity coefficient to the Doppler reactivity coefficient more than 10x10 3 kg·cm 3 which is two to three times larger than that at beginning of cycle (BOC) in an ordinary PWR and realized by elimination of the chemical shim, the 12% lower average linear heat generation rate of 17.9 kW/m, and the 1.5 times larger pressurizer volume than those of the ordinary PWR. (author)

  15. Simplified analysis of passive residual heat removal systems for small size PWR's

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1992-02-01

    The function and general objectives of a passive residual heat removal system for small size PWR's are defined. The characteristic configuration, the components and the operation modes of this system are concisely described. A preliminary conceptual specification of this system, for a small size PWR of 400 MW thermal, is made analogous to the decay heat removal system of the AP-600 reactor. It is shown by analytic models that such passive systems can dissipate 2% of nominal power within the thermal limits allowed to the reactor fuel elements. (author)

  16. Large scale experiments with a 5 MW sodium/air heat exchanger for decay heat removal

    International Nuclear Information System (INIS)

    Stehle, H.; Damm, G.; Jansing, W.

    1994-01-01

    Sodium experiments in the large scale test facility ILONA were performed to demonstrate proper operation of a passive decay heat removal system for LMFBRs based on pure natural convection flow. Temperature and flow distributions on the sodium and the air side of a 5 MW sodium/air heat exchanger in a natural draught stack were measured during steady state and transient operation in good agreement with calculations using a two dimensional computer code ATTICA/DIANA. (orig.)

  17. Behavior study on Na heat pipe in passive heat removal system of new concept molten salt reactor

    International Nuclear Information System (INIS)

    Wang Chenglong; Tian Wenxi; Su Guanghui; Zhang Dalin; Wu Yingwei; Qiu Suizheng

    2013-01-01

    The high temperature Na heat pipe is an effective device for transporting heat, which is characterized by remarkable advantages in conductivity, isothermally and passively working. The application of Na heat pipe on passive heat removal system of new concept molten salt reactor (MSR) is significant. The transient performance of high temperature Na heat pipe was simulated by numerical method under the MSR accident. The model of the Na heat pipe was composed of three conjugate heat transfer zones, i.e. the vapor, wick and wall. Based on finite element method, the governing equations were solved by making use of FORTRAN to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results show that the high temperature Na heat pipe has a good performance on operating characteristics and high heat transfer efficiency from the frozen state. (authors)

  18. Technique for the Laparoscopic Removal of Essure Microinserts.

    Science.gov (United States)

    Mahmoud, Mohamad S

    2016-01-01

    To describe our technique for the laparoscopic removal of Essure microinserts (Bayer HealthCare Pharmaceuticals Inc., Whippany, NJ). Step-by-step explanation of the procedure using video (Canadian Task Force classification III). Hysteroscopic sterilization using tubal microinsert devices has generally been reported to be well tolerated in terms of procedure-related pain. Persistent pelvic pain requiring microinsert removal has been described in a few case reports and series and was estimated at 0.16% of cases (7 cases [49/4,274]) in a large retrospective study. Removal is usually performed at the patient's request and/or because of persistent pelvic pain unresponsive to other treatments with no other etiologies found. In general, the pain starts at the initial insertion and persists thereafter. Both laparoscopic and hysteroscopic removal approaches have been described in the few cases reported. In this video, we describe our technique for the surgical management of pelvic pain resulting from Essure microinserts. We performed laparoscopic removal of bilateral Essure microinserts in a 30-year-old G3P3 (Gravida 3 Para 3) with bilateral Essure devices placed 2 years before the procedure; hysterosalpingogram confirmed appropriate placement. The patient was suffering from bilateral sharp pelvic pain since insertion that was related to positional change and movements but unrelated to periods along with menorrhagia. A pelvic ultrasound showed a small intramural uterine leiomyoma. She failed a trial of treatment of her symptoms with a levonorgestrel intrauterine device. The patient requested removal of her Essure microinserts and endometrial ablation. She underwent laparoscopic bilateral Essure microinsert removal and bilateral salpingectomy along with hysteroscopic removal of the levonorgestrel intrauterine device and endometrial ablation. Her surgery was uneventful, and she was discharged the day of the surgery. Her symptoms resolved completely after the procedure

  19. Characterizing convective heat transfer using infrared thermography and the heated-thin-foil technique

    International Nuclear Information System (INIS)

    Stafford, Jason; Walsh, Ed; Egan, Vanessa

    2009-01-01

    Convective heat transfer, due to axial flow fans impinging air onto a heated flat plate, is investigated with infrared thermography to assess the heated-thin-foil technique commonly used to quantify two-dimensional heat transfer performance. Flow conditions generating complex thermal profiles have been considered in the analysis to account for dominant sources of error in the technique. Uncertainties were obtained in the measured variables and the influences on the resultant heat transfer data are outlined. Correction methods to accurately account for secondary heat transfer mechanisms were developed and results show that as convective heat transfer coefficients and length scales decrease, the importance of accounting for errors increases. Combined with flow patterns that produce large temperature gradients, the influence of heat flow within the foil on the resultant heat transfer becomes significant. Substantial errors in the heat transfer coefficient are apparent by neglecting corrections to the measured data for the cases examined. Methods to account for these errors are presented here, and demonstrated to result in an accurate measurement of the local heat transfer map on the surface

  20. PBMR spent fuel bulk dry storage heat removal - HTR2008-58170

    International Nuclear Information System (INIS)

    De Wet, G. J.; Dent, C.

    2008-01-01

    A low decay heat (implying Spent Fuel (SF) pebbles older than 8-9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks' vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading/unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell. (authors)

  1. Heat Transfer and Cooling Techniques at Low Temperature

    CERN Document Server

    Baudouy, B

    2014-07-17

    The first part of this chapter gives an introduction to heat transfer and cooling techniques at low temperature. We review the fundamental laws of heat transfer (conduction, convection and radiation) and give useful data specific to cryogenic conditions (thermal contact resistance, total emissivity of materials and heat transfer correlation in forced or boiling flow for example) used in the design of cooling systems. In the second part, we review the main cooling techniques at low temperature, with or without cryogen, from the simplest ones (bath cooling) to the ones involving the use of cryocoolers without forgetting the cooling flow techniques.

  2. Heat Transfer and Cooling Techniques at Low Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Baudouy, B [Saclay (France)

    2014-07-01

    The first part of this chapter gives an introduction to heat transfer and cooling techniques at low temperature. We review the fundamental laws of heat transfer (conduction, convection and radiation) and give useful data specific to cryogenic conditions (thermal contact resistance, total emissivity of materials and heat transfer correlation in forced or boiling flow for example) used in the design of cooling systems. In the second part, we review the main cooling techniques at low temperature, with or without cryogen, from the simplest ones (bath cooling) to the ones involving the use of cryocoolers without forgetting the cooling flow techniques.

  3. Photovoltaic cell electrical heating system for removing snow on panel including verification.

    Science.gov (United States)

    Weiss, Agnes; Weiss, Helmut

    2017-11-16

    Small photovoltaic plants in private ownership are typically rated at 5 kW (peak). The panels are mounted on roofs at a decline angle of 20° to 45°. In winter time, a dense layer of snow at a width of e.g., 10 cm keeps off solar radiation from the photovoltaic cells for weeks under continental climate conditions. Practically, no energy is produced over the time of snow coverage. Only until outside air temperature has risen high enough for a rather long-time interval to allow partial melting of snow; the snow layer rushes down in an avalanche. Following this proposal, snow removal can be arranged electrically at an extremely positive energy balance in a fast way. A photovoltaic cell is a large junction area diode inside with a threshold voltage of about 0.6 to 0.7 V (depending on temperature). This forward voltage drop created by an externally driven current through the modules can be efficiently used to provide well-distributed heat dissipation at the cell and further on at the glass surface of the whole panel. The adhesion of snow on glass is widely reduced through this heating in case a thin water film can be produced by this external short time heating. Laboratory experiments provided a temperature increase through rated panel current of more than 10 °C within about 10 min. This heating can initiate the avalanche for snow removal on intention as described before provided the clamping effect on snow at the edge of the panel frame is overcome by an additional heating foil. Basics of internal cell heat production, heating thermal effects in time course, thermographic measurements on temperature distribution, power circuit opportunities including battery storage elements and snow-removal under practical conditions are described.

  4. Removal of corrosion products of construction materials in heat carrier

    International Nuclear Information System (INIS)

    1975-01-01

    A review of reported data has been made on the removal of structural material corrosion products into the heat-carrying agent of power reactors. The corrosion rate, and at the same time, removal of corrosion products into the heat-carrying agent (water) decreases with time. Thus, for example, the corrosion rate of carbon steel in boiling water at 250 deg C and O 2 concentration of 0.1 mg/1 after 3000 hr is 0.083 g/m 2 . day; after 9000 hr the corrosion rate has been reduced 2.5 times. Under static conditions the transfer rate of corrosion products into water has been smaller than in the stream and also depends on time. The corrosion rate of carbon steel under nuclear plant operating conditions is almost an order higher over that of steel Kh18N10T

  5. Design of a natural draft air-cooled condenser and its heat transfer characteristics in the passive residual heat removal system for 10 MW molten salt reactor experiment

    International Nuclear Information System (INIS)

    Zhao, Hangbin; Yan, Changqi; Sun, Licheng; Zhao, Kaibin; Fa, Dan

    2015-01-01

    As one of the Generation IV reactors, Molten Salt Reactor (MSR) has its superiorities in satisfying the requirements on safety. In order to improve its inherent safety, a concept of passive residual heat removal system (PRHRS) for the 10 MW Molten Salt Reactor Experiment (MSRE) was put forward, which mainly consisted of a fuel drain tank, a feed water tank and a natural draft air-cooled condenser (NDACC). Besides, several valves and pipes are also included in the PRHRS. A NDACC for the PRHRS was preliminarily designed in this paper, which contained a finned tube bundle and a chimney. The tube bundle was installed at the bottom of the chimney for increasing the velocity of the air across the bundle. The heat transfer characteristics of the NDACC were investigated by developing a model of the PRHRS using C++ code. The effects of the environmental temperature, finned tube number and chimney height on heat removal capacity of the NDACC were analyzed. The results show that it has sufficient heat removal capacity to meet the requirements of the residual heat removal for MSRE. The effects of these three factors are obvious. With the decay heat reducing, the heat dissipation power declines after a short-time rise in the beginning. The operation of the NDACC is completely automatic without the need of any external power, resulting in a high safety and reliability of the reactor, especially once the accident of power lost occurs to the power plant. - Highlights: • A model to study the heat transfer characteristics of the NDACC was developed. • The NDACC had sufficient heat removal capacity to remove the decay heat of MSRE. • NDACC heat dissipation power depends on outside temperature and condenser geometry. • As time grown, the effects of outside temperature and condenser geometry diminish. • The NDACC could automatically adjust its heat removal capacity

  6. Natural convection as the way of heat removal from fast reactor core at cooldown regimes

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Kuzina, J.A.; Uhov, V.A.; Sorokin, G.A.

    2000-01-01

    The problems of thermohydraulics in fast reactors at cooldown regimes at heat removal by natural convection are considered The results of experiments and calculations obtained in various countries in this area are presented. The special attention is given to heat removal through inter-assembly space in the core and also to problems of thermohydraulics in the upper plenum. (author)

  7. Post-accident heat removal ''information exchange''

    International Nuclear Information System (INIS)

    Plein, H.G.; Carlson, G.A.

    1975-01-01

    The in-core molten pool experiments are designed to produce a pool of fission heated temperature and flow patterns of such pools, and evaluate the barrier melt-through potential of the molten UO 2 . The first experiments, to be conducted this fiscal year in the Annular Core Pulse Reactor, will be uncomplicated and multiply-contained to prove containment design and to provide initial information on fission heated molten pool characteristics. Concurrent with the in-core experiments, high temperature ultrasonic techniques are being developed to measure UO 2 temperatures up to and above the melting point for use in later more definitive experiments scheduled for FY77

  8. Valve arrangement for a nuclear plant residual heat removal system

    International Nuclear Information System (INIS)

    Fidler, G.L.; Hill, R.A.; Carrera, J.P.

    1978-01-01

    Disclosed is an improved valve arrangement for a two-train Residual Heat Removal System (RHRS) of a nuclear reactor plant which ensures operational integrity of the system under single failure circumstances including loss of one of two electrical power sources

  9. Numerical analysis of high-power broad-area laser diode with improved heat sinking structure using epitaxial liftoff technique

    Science.gov (United States)

    Kim, Younghyun; Sung, Yunsu; Yang, Jung-Tack; Choi, Woo-Young

    2018-02-01

    The characteristics of high-power broad-area laser diodes with the improved heat sinking structure are numerically analyzed by a technology computer-aided design based self-consistent electro-thermal-optical simulation. The high-power laser diodes consist of a separate confinement heterostructure of a compressively strained InGaAsP quantum well and GaInP optical cavity layers, and a 100-μm-wide rib and a 2000-μm long cavity. In order to overcome the performance deteriorations of high-power laser diodes caused by self-heating such as thermal rollover and thermal blooming, we propose the high-power broad-area laser diode with improved heat-sinking structure, which another effective heat-sinking path toward the substrate side is added by removing a bulk substrate. It is possible to obtain by removing a 400-μm-thick GaAs substrate with an AlAs sacrificial layer utilizing well-known epitaxial liftoff techniques. In this study, we present the performance improvement of the high-power laser diode with the heat-sinking structure by suppressing thermal effects. It is found that the lateral far-field angle as well as quantum well temperature is expected to be improved by the proposed heat-sinking structure which is required for high beam quality and optical output power, respectively.

  10. RCS pressure under reduced inventory conditions following a loss of residual heat removal

    International Nuclear Information System (INIS)

    Palmrose, D.E.; Hughes, E.D.; Johnsen, G.W.

    1992-01-01

    The thermal-hydraulic response of a closed-reactor coolant system to loss of residual heat removal (RHR) cooling is investigated. The processes examined include: core coolant boiling and steam generator reflux condensation, pressure increase on the primary side, heat transfer mechanisms on the steam generator primary and secondary sides, and effects of noncondensible gas on heat transfer processes

  11. Nuclear power plant with improved arrangements for the removal of post fission and emergency heating

    International Nuclear Information System (INIS)

    Buescher, E.; Vinzens, K.

    1977-01-01

    This is concerned with additional equipment for emergency heat removal in a sodium cooled reactor, which operates on failure of the post fission heat removal system. The space for pressure relieving spaces and concrete masses as heat sinks within the reactor cell is no longer required. In this nuclear power plant, a heat exchanger chain transmits heat and power: There is a first sodium circuit between pressure vessel and the first heat exchanger, a second one between the first and second heat excahngers, and a third (Steam) circuit with turbine, condenser and return pump. A fourth circuit connects the secondary side of the condenser with a cooling tower. There is a threee component heat excahgner in the primary circuit after the first heat exchanger, which is built around the first heat exchanger, and is sealed into an unloading space. This space is situated next to the reactor cell and is above the operating level of the sodium in the pressure vessel. It is connected to the cell by an upper duct, normally closed by a bursting disc, and by a lower duct. In the three comopnent heat exchanger, a liquid lead-bismuth eutectic mixture transmits the heat from sodium pipes to water pipes. In normal operation it is used for steam superheating or feedwater preheating. The three component heat exchanger bridges the first and second heat exchangers as an emergency heat exchanger. If in such a case the post fission heat removal has failed, the sodium evaporating in the pressure vessel flows into the unloading space and condenses on the ribs of the emergency heat exchanger. The post fission heat is fed by the water secondary medium directly into the tertiary circuit. The sodium condensate flows back from the unloading space via the lower duct into the reactor cell and maintains the emergency level there. (RW) 891 RW [de

  12. A decay heat removal methodology for reuseable orbital transfer vehicles

    Science.gov (United States)

    McDaniel, Patrick J.; Perkins, David R.

    1992-07-01

    Operation of a nuclear thermal rocket(NTR) as the propulsion system for a reusable orbital transfer vehicle has been considered. This application is the most demanding in terms of designing a multiple restart capability for an NTR. The requirements on a NTR cooling system associated with the nuclear decay heat stored during operation have been evaluated, specifically for a Particle Bed Reactor(PBR) configuration. A three mode method of operation has been identified as required to adequately remove the nuclear decay heat.

  13. Efficient on-chip hotspot removal combined solution of thermoelectric cooler and mini-channel heat sink

    International Nuclear Information System (INIS)

    Hao, Xiaohong; Peng, Bei; Xie, Gongnan; Chen, Yi

    2016-01-01

    Highlights: • A combined solution of thermoelectric cooler (TEC) and mini-channel heat sink to remove the hotspot of the chip has been proposed. • The TEC's mathematical model is established to assess its work performance. • A comparative study on the proposed efficient On-Chip Hotspot Removal Combined Solution. - Abstract: Hotspot will significantly degrade the reliability and performance of the electronic equipment. The efficient removal of hotspot can make the temperature distribution uniform, and ensure the reliable operation of the electronic equipment. This study proposes a combined solution of thermoelectric cooler (TEC) and mini-channel heat sink to remove the hotspot of the chip in the electronic equipment. Firstly, The TEC's mathematical model is established to assess its work performance under different boundary conditions. Then, the hotspot removal capability of the TEC is discussed for different cooling conditions, which has shown that the combined equipment has better hotspot removal capability compared with others. Finally, A TEC is employed to investigate the hotspot removal capacity of the combined solution, and the results have indicated that it can effectively remove hotspot in the diameter of 0.5 mm, the power density of 600W/cm 2 when its working current is 3A and heat transfer thermal resistance is 0 K/W.

  14. Optimized design of an ex-vessel cooling thermosyphon for decay heat removal in SFR

    International Nuclear Information System (INIS)

    Choi, Jae Young; Jeong, Yong Hoon; Song, Sub Lee; Chang, Soon Heung

    2017-01-01

    Passive decay heat removal and sodium fire are two major key issues of nuclear safety in sodium-cooled fast reactor (SFR). Several decay heat removal systems (DHR) were suggested for SFR around the world so far. Those DHRS mainly classified into two concepts: Direct reactor cooling system and ex-vessel cooling system. Direct reactor cooling method represented by PDHRS from PGSFR has disadvantages on its additional in-vessel structure and potential sodium fire risk due to the sodium-filled heat exchanger exposed to air. Contrastively, ex-vessel cooling method represented by RVACS from PRISM has low decay heat removal performance, which cannot be applicable to large scale reactors, generally over 1000 MWth. No passive DHRSs which can solve both side of disadvantages has been suggested yet. The goal of this study was to propose ex-vessel cooling system using two-phase closed thermosyphon to compensate the disadvantages of the past DHRSs. Reference reactor was Innovative SFR (iSFR), a pool-type SFR designed by KAIST and featured by extended core lifetime and increased thermal efficiency. Proposed ex-vessel cooling system consisted of 4 trains of thermosyphons and designed to remove 1% of thermal power with 10% of margin. The scopes of this study were design of proposed passive DHRS, validation of system analysis and optimization of system design. Mercury was selected as working fluid to design ex-vessel thermosyphon in consideration of system geometry, operating temperature and required heat flux. SUS 316 with chrome coated liner was selected as case material to resist against high corrosivity of mercury. Thermosyphon evaporator was covered on the surface of reactor vessel as the geometry of hollow shell filled with mercury. Condenser was consisted of finned tube bundles and was located in isolated water pool, the ultimate heat sink. Operation limits and thermal resistance was estimated to guarantee whether the design was adequate. System analysis was conducted by in

  15. A PRA case study of extended long term decay heat removal for shutdown risk assessment

    International Nuclear Information System (INIS)

    Roglans, J.; Ragland, W.A.; Hill, D.J.

    1992-01-01

    A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL). The results of this PRA have shown that the decay heat removal system for EBR-II is extremely robust and reliable. In addition, the methodology used demonstrates how the actions of other systems not normally used for actions of other systems not normally used for decay heat removal can be used to expand the mission time of the decay heat removal system and further increase its reliability. The methodology may also be extended to account for the impact of non-safety systems in enhancing the reliability of other dedicated safety systems

  16. RELAP5 and SIMMER-III code assessment on CIRCE decay heat removal experiments

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Polidori, Massimiliano; Meloni, Paride; Tarantino, Mariano; Di Piazza, Ivan

    2015-01-01

    Highlights: • The CIRCE DHR experiments simulate LOHS+LOF transients in LFR systems. • Decay heat removal by natural circulation through immersed heat exchangers is investigated. • The RELAP5 simulation of DHR experiments is presented. • The SIMMER-III simulation of DHR experiments is presented. • The focus is on the transition from forced to natural convection and stratification in a large pool. - Abstract: In the frame of THINS Project of the 7th Framework EU Program on Nuclear Fission Safety, some experiments were carried out on the large scale LBE-cooled CIRCE facility at the ENEA/Brasimone Research Center to investigate relevant safety aspects associated with the removal of decay heat through heat exchangers (HXs) immersed in the primary circuit of a pool-type lead fast reactor (LFR), under loss of heat sink (LOHS) accidental conditions. The start-up and operation of this decay heat removal (DHR) system relies on natural convection on the primary side and then might be affected by coolant mixing and temperature stratification phenomena occurring in the LBE pool. The main objectives of the CIRCE experimental campaign were to verify the behavior of the DHR system under representative accidental conditions and provide a valuable database for the assessment of both CFD and system codes. The reproduced accidental conditions refer to a station blackout scenario, namely a protected LOHS and loss of flow (LOF) transient. In this paper the results of 1D RELAP5 and 2D SIMMER-III simulations are compared with the experimental data of more representative DHR transients T-4 and T-5 in order to verify the capability of these codes to reproduce both forced and natural convection conditions observed in the primary circuit and the right operation of the DHR system for decay heat removal. Both codes are able to reproduce the stationary conditions and with some uncertainties the transition to natural convection conditions until the end of the transient phase. The trend

  17. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    Science.gov (United States)

    Missirlian, M.; Richou, M.; Riccardi, B.; Gavila, P.; Loarer, T.; Constans, S.

    2011-12-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m-2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m-2 for the CFC-armoured tiles and 15 MW m-2 for the W-armoured tiles, respectively.

  18. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M; Richou, M; Loarer, T; Riccardi, B; Gavila, P; Constans, S

    2011-01-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m - 2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m - 2 for the CFC-armoured tiles and 15 MW m - 2 for the W-armoured tiles, respectively.

  19. Preliminary study of the decay heat removal strategy for the gas demonstrator allegro

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, Gusztáv, E-mail: gusztav.mayer@energia.mta.hu [Hungarian Academy of Sciences, Centre for Energy Research, P.O. Box 49, H-1525 Budapest (Hungary); Bentivoglio, Fabrice, E-mail: fabrice.bentivoglio@cea.fr [CEA/DEN/DM2S/STMF/LMES, F-38054, Grenoble (France)

    2015-05-15

    Highlights: • Improved decay heat removal strategy was adapted for the 75 MW ALLEGRO MOX core. • New nitrogen injection strategy was proposed for the DEC LOCA transients. • Preliminary CATHARE study shows that most of the investigated transients fulfill criteria. • Further improvements and optimizations are needed for nitrogen injection. - Abstract: The helium cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected in the frame of the Generation IV International Forum. Since no gas cooled fast reactor has ever been built, a medium power demonstrator reactor – named ALLEGRO – is necessary on the road towards the 2400 MWth GFR power reactor. The French Commissariat à l’Energie Atomique (CEA) completed a wide range of studies during the early stage of development of ALLEGRO, and later the ALLEGRO reactor concept was developed in several European Union projects in parallel with the GFR2400. The 75 MW thermal power ALLEGRO is currently developed in the frame of the European ALLIANCE project. As a result of the collaboration between CEA and the Hungarian Academy of Sciences Centre for Energy Research (MTA EK) new improvements were done in the safety approach of ALLEGRO. A complete Decay Heat Removal (DHR) strategy was devised, relying on the primary circuits as a first way to remove decay heat using pony-motors to drive the primary blowers, and on the secondary and tertiary circuits being able to work in forced or natural circulation. Three identical dedicated loops circulating in forced convection are used as a second way to remove decay heat, and these loops can circulate in natural convection for pressurized transients, providing a third way to remove decay heat in case of accidents when the primary circuit is still under pressure. The possibility to use nitrogen to enhance both forced and natural circulation is discussed. This DHR strategy is supported by a wide range of accident transient simulations performed using the CATHARE2 code

  20. Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

    Energy Technology Data Exchange (ETDEWEB)

    Murav’ev, V. P., E-mail: murval1@mail.ru

    2016-07-15

    The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

  1. Techniques for removing contaminants from optical surfaces

    International Nuclear Information System (INIS)

    Stowers, I.F.; Patton, H.G.

    1978-01-01

    Particle removal procedures such as plasma cleaning, ultrasonic agitation of solvents, detergents, solvent wiping, mild abrasives, vapor degreasing, high pressure solvent spraying and others have been evaluated and the results are reported here. Wiping with a lens tissue wetted with an organic solvent and high pressure fluid spraying are the only methods by which particles as small as 5 μm can be effectively removed. All of the other methods tested were found to be at least two orders of magnitude less effective at removing small insoluble particles. An additional and as yet unresolved problem is the development of a reliable method for evaluating particulate surface cleanliness. Without such a reproducible monitoring technique, the large diversity of cleaning methods currently available cannot be quantitatively evaluated

  2. Latest innovations for tattoo and permanent makeup removal.

    Science.gov (United States)

    Mao, Johnny C; DeJoseph, Louis M

    2012-05-01

    The goal of this article is to reveal the latest techniques and advances in laser removal of both amateur and professional tattoos, as well as cosmetic tattoos and permanent makeup. Each pose different challenges to the removing physician, but the goal is always the same: removal without sequelae. The authors' technique is detailed, and discussion of basic principles of light reflection, ink properties, effects of laser energy and heat, and outcomes and complications of tattoo removal are presented. Copyright © 2012 Elsevier Inc. All rights reserved.

  3. Passive safety systems for decay heat removal of MRX

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, M; Iida, H; Hoshi, T [Japan Atomic Energy Research Inst., Ibaraki (Japan). Nuclear Ship System Lab.

    1996-12-01

    The MRX (marine Reactor X) is an advanced marine reactor, its design has been studied in Japan Atomic Energy Research Institute. It is characterized by four features, integral type PWR, in-vessel type control rod drive mechanisms, water-filled containment vessel and passive decay heat removal system. A water-filled containment vessel is of great advantage since it ensures compactness of a reactor plant by realizing compact radiation shielding. The containment vessel also yields passive safety of MRX in the event of a LOCA by passively maintaining core flooding without any emergency water injection. Natural circulation of water in the vessels (reactor and containment vessels) is one of key factors of passive decay heat removal systems of MRX, since decay heat is transferred from fuel rods to atmosphere by natural circulation of the primary water, water in the containment vessel and thermal medium in heat pipe system for the containment vessel water cooling in case of long terms cooling after a LOCA as well as after reactor scram. Thus, the ideal of water-filled containment vessel is considered to be very profitable and significant in safety and economical point of view. This idea is, however, not so familiar for a conventional nuclear system, so experimental and analytical efforts are carried out for evaluation of hydrothermal behaviours in the reactor pressure vessel and in the containment vessel in the event of a LOCA. The results show the effectiveness of the new design concept. Additional work will also be conducted to investigate the practical maintenance of instruments in the containment vessel. (author). 4 refs, 9 figs, 2 tabs.

  4. Application of grey model on analyzing the passive natural circulation residual heat removal system of HTR-10

    Institute of Scientific and Technical Information of China (English)

    ZHOU Tao; PENG Changhong; WANG Zenghui; WANG Ruosu

    2008-01-01

    Using the grey correlation analysis, it can be concluded that the reactor pressure vessel wall temperature has the strongest effect on the passive residual heat removal system in HTR (High Temperature gas-cooled Reactor),the chimney height takes the second place, and the influence of inlet air temperature of the chimney is the least. This conclusion is the same as that analyzed by the traditional method. According to the grey model theory, the GM(1,1) and GM(1, 3) model are built based on the inlet air temperature of chimney, pressure vessel temperature and the chimney height. Then the effect of three factors on the heat removal power is studied in this paper. The model plays an important role on data prediction, and is a new method for studying the heat removal power. The method can provide a new theoretical analysis to the passive residual heat removal system of HTR.

  5. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Lap-Yan, C.; Wie, T. Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  6. An estimation of core damage frequency of a pressurized water reactor during midloop operation due to loss of residual heat removal

    International Nuclear Information System (INIS)

    Chao, C.C.; Chen, C.T.; Lee, M.

    1995-01-01

    The core damage frequency caused by loss of residual heat removal (RHR) events was assessed during midloop operation of a Westinghouse-designed three-loop pressurized water reactor. The assessment considers two types of outages (refueling and drained maintenance) and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events were identified and human error probabilities were quantified using the human cognitive reliability (HCR) and the technique for human error rate prediction (THERP) models. The results showed that the core damage frequency caused by loss of RHR events during midloop operation was 3.4 x 10 -5 per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering midloop operation. The establishment of reflux cooling, i.e., decay heat removal through the steam generator secondary side, also plays an important role in mitigating the loss of RHR events during midloop operation

  7. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely.

  8. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    International Nuclear Information System (INIS)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu

    2016-01-01

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely

  9. Heat Removal Performance of Hybrid Control Rod for Passive In-Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    The two-phase closed heat transfer device can be divided by thermosyphon heat pipe and capillary wicked heat pipe which uses gravitational force or capillary pumping pressure as a driving force of the convection of working fluid. If there is a temperature difference between reactor core and ultimate heat sink, the decay heat removal and reactor shutdown is possible at any accident conditions without external power sources. To apply the hybrid control rod to the commercial nuclear power plants, its modelling about various parameters is the most important work. Also, its unique geometry is coexistence of neutron absorber material and working fluid in a cladding material having annular vapor path. Although thermosyphon heat pipe (THP) or wicked heat pipe (WHP) shows high heat transfer coefficients for limited space, the maximum heat removal capacity is restricted by several phenomena due to their unique heat transfer mechanism. Validation of the existing correlations on the annular vapor path thermosyphon (ATHP) which has different wetted perimeter and heated diameter must be conducted. The effect of inner structure, and fill ratio of the working fluid on the thermal performance of heat pipe has not been investigated. As a first step of the development of hybrid heat pipe, the ATHP which contains neutron absorber in the concentric thermosyphon (CTHP) was prepared and the thermal performance of the annular thermosyphon was experimentally studied. The heat transfer characteristics and flooding limit of the annular vapor path thermosyphon was studied experimentally to model the performance of hybrid control rod. The following results were obtained: (1) The annular vapor path thermosyphon showed better evaporation heat transfer due to the enhanced convection between adiabatic and condenser section. (2) Effect of fill ratio on the heat transfer characteristics was negligible. (3) Existing correlations about flooding limit of thermosyphon could not reflect the annular vapor

  10. Analysis and testing of W-DHR system for decay heat removal in the lead-cooled ELSY reactor

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Meloni, Paride; Polidori, Massimiliano; Gaggini, Piero; Labanti, Valerio; Tarantino, Mariano; Cinotti, Luciano; Presciuttini, Leonardo

    2009-01-01

    An innovative LFR system that complies with GEN IV goals is under design in the frame of ELSY European project. ELSY is a lead-cooled pool-type reactor of about 1500 MW thermal power which normally relies on the secondary system for decay heat removal. Since the secondary system is not safety-grade and must be fully depressurized in case of detection of a steam generator tube rupture, an independent and much reliable decay heat removal (DHR) system is foreseen on the primary side. Owing to the limited capability of the Reactor Vessel Air Cooling System (RVACS) in this large power reactor, additional safety-grade loops equipped with coolers immersed in the primary coolant are necessary for an efficient removal of decay heat. Some of these loops (W-DHR) are of innovative design and may operate with water at atmospheric pressure. In the frame of the ICE program to be performed on the integral facility CIRCE at ENEA/Brasimone research centre within the EUROTRANS European project, integral circulation experiments with core heat transport and heat removal by steam generator will be conducted in a reactor pool-type configuration. Taking advantage from this experimental program, a mock-up of W-DHR heat exchanger will be tested in order to investigate its functional behavior for decay heat removal. Some pre-test calculations of W-DHR heat exchanger operation in CIRCE have been performed with the RELAP5 thermal-hydraulic code in order to support the heat exchanger design and test conduct. In this paper the experimental activity to be conducted in CIRCE and main results from W-DHR pre-test calculations are presented, along with a preliminary investigation of the W-DHR system efficiency in ELSY configuration. (author)

  11. Performance of ALMR passive decay heat removal system

    International Nuclear Information System (INIS)

    Boardman, C.E.; Hunsbedt, A.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the small (471 MWt) modular reactor to the environmental air by natural convection heat transfer. The system has no active components, requires no operator action to initiate, and is inherently reliable. The RVACS can perform its function under off-normal or degraded operating conditions without significant loss in performance. Several such events are described and the RVACS thermal performance for each is given and compared to the normal operation performance. The basic RVACS performance as well as the performance during several off-normal events have been updated to reflect design changes for recycled fuel with minor actinides for end of equilibrium cycle conditions. The performance results for several other off-normal events involving various degrees of RVACS air flow passage blockages are presented. The results demonstrated that the RVACS is unusually tolerant to a wide range of postulated faults. (author)

  12. Heat Removal from Bipolar Transistor by Loop Heat Pipe with Nickel and Copper Porous Structures

    Science.gov (United States)

    Smitka, Martin; Malcho, Milan

    2014-01-01

    Loop heat pipes (LHPs) are used in many branches of industry, mainly for cooling of electrical elements and systems. The loop heat pipe is a vapour-liquid phase-change device that transfers heat from evaporator to condenser. One of the most important parts of the LHP is the porous wick structure. The wick structure provides capillary force to circulate the working fluid. To achieve good thermal performance of LHP, capillary wicks with high permeability and porosity and fine pore radius are expected. The aim of this work was to develop porous structures from copper and nickel powder with different grain sizes. For experiment copper powder with grain size of 50 and 100 μm and nickel powder with grain size of 10 and 25 μm were used. Analysis of these porous structures and LHP design are described in the paper. And the measurements' influences of porous structures in LHP on heat removal from the insulated gate bipolar transistor (IGBT) have been made. PMID:24959622

  13. Overview report of RAMONA-NEPTUN program on passive decay heat removal

    International Nuclear Information System (INIS)

    Weinberg, D.; Rust, K.; Hoffmann, H.

    1996-03-01

    The design of the advanced sodium-cooled European Fast Reactor provides a safety graded decay heat removal concept which ensures the coolability of the primary system by natural convection when forced cooling is lost. The findings of the RAMONA and NEPTUN experiments indicate that the decay heat can be safely removed by natural convection. The operation of the decay heat exchangers being installed in the upper plenum causes the formation of a thermal stratification associated with a pronounced temperature gradient. The vertical extent of the stratification and the qualitity of the gradient are depending on the fact whether a permeable or an impermeable shell covers the above core structure. A delayed startup time of the decay heat exchangers leads only to a slight increase of the temperatures in the upper plenum. A complete failure of half of the decay heat exchangers causes a higher temperature level in the primary system, but does not alter the global temperature distribution. The transient development of the temperatures is faster going on in a three-loop model than in a four-loop model due to the lower amount of heat stored in the compacter primary vessel. If no coolant reaches the core inlet side via the intermediate heat exchangers, the core remains coolable. In this case, cold water of the upper plenum penetrates into the subassemblies (thermosyphon effects) and the interwrapper spaces existing in the NEPTUN core. The core coolability from above is feasible without any difficulty though the temperatures increase to a minor degree at the top end of the core. The thermal hydraulic computer code FLUTAN was applied for the 3D numerical simulation of the majority of the steady state RAMONA and NEPTUN tests as well as for selected transient RAMONA tests. (orig./HP) [de

  14. Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for the Advanced Integral Type Reactor

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki; Lee, Un-Chul

    2004-01-01

    Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable the natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with the operation of the PRHRS. (authors)

  15. [Technique for removing donor sclera by eyeball extrusion].

    Science.gov (United States)

    González Del Valle, F; Álvarez Portela, M; Lara Medina, J; Celis Sánchez, J; Barrajón Rodríguez, A

    2012-09-01

    To describe a surgery technique for removing donor sclera tissue after corneo-scleral button excision. The extrusion technique is easy to perform. It allows the complete scleral extraction its total clean up to be performed, as well as making easier to isolate the retina and uveal tissue. This technique could have an important role in the anatomical and morphological study of ocular structures. Copyright © 2011 Sociedad Española de Oftalmología. Published by Elsevier Espana. All rights reserved.

  16. Concepts for passive heat removal and filtration systems under core meltdown conditions

    International Nuclear Information System (INIS)

    Wilhelm, J.G.; Neitzel, H.-J.

    1993-01-01

    The objective of the new containment concept being developed by KfK is the complete passive enclosure of a power reactor after a core meltdown accident by means of a solid containment structure and passive removal of the decay heat. This is to be accomplished by cooling the containment walls with ambient air, with thermoconvection as the driving force. The concept of the containment is described. Data are given of the heat removal and the requirements for filtration of the exhaust air, which is contaminated due to the leak rate assumed for the inner containment. The concept for the filter system is described. Various solutions for reduction of the large volumetric flow to be filtered are discussed. 3 refs., 8 figs

  17. Analysis of Decay Heat Removal by Natural Convection in LMR with a Combined Steam Generator

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Eoh, Jae Hyuk; Han, Ji Woong; Lee, Tae Ho

    2011-01-01

    Liquid metal reactors (LMRs) conventionally employ an intermediate heat transport system (IHTS) to protect the nuclear core during a sodium-water reaction (SWR) event. However these SWR-related components increase plant construction costs. In order to eliminate the need for an IHTS, a combined steam generator, which is an integrated heat exchanger of a steam generator and intermediate heat exchanger (IHX), was proposed by the Korea Atomic Energy Research Institute (KAERI). The objective of this work is to analyze the natural circulation heat removal capability of the rector system using a combined steam generator. As a means of decay heat removal, a normal heat transport path is composed of a primary sodium system, intermediate lead-bismuth circuit combined with SG and steam/water system. This paper presents the results of the possible temperature and natural circulation flows in all circuits during a steady state for a given reactor power level varied as a function of time

  18. Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

    Energy Technology Data Exchange (ETDEWEB)

    Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-05-15

    Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

  19. Comparison of residual NAPL source removal techniques in 3D metric scale experiments

    Science.gov (United States)

    Atteia, O.; Jousse, F.; Cohen, G.; Höhener, P.

    2017-07-01

    This study compared four treatment techniques for the removal of a toluene/n-decane as NAPL (Non Aqueous Phase Liquid) phase mixture in identical 1 cubic meter tanks filled with different kind of sand. These four treatment techniques were: oxidation with persulfate, surfactant washing with Tween80®, sparging with air followed by ozone, and thermal treatment at 80 °C. The sources were made with three lenses of 26 × 26 × 6.5 cm, one having a hydraulic conductivity similar to the whole tank and the two others a value 10 times smaller. The four techniques were studied after conditioning the tanks with tap water during approximately 80 days. The persulfate treatment tests showed average removal of the contaminants but significant flux decrease if density effects are considered. Surfactant flushing did not show a highly significant increase of the flux of toluene but allowed an increased removal rate that could lead to an almost complete removal with longer treatment time. Sparging removed a significant amount but suggests that air was passing through localized gas channels and that the removal was stagnating after removing half of the contamination. Thermal treatment reached 100% removal after the target temperature of 80 °C was kept during more than 10 d. The experiments emphasized the generation of a high-spatial heterogeneity in NAPL content. For all the treatments the overall removal was similar for both n-decane and toluene, suggesting that toluene was removed rapidly and n-decane more slowly in some zones, while no removal existed in other zones. The oxidation and surfactant results were also analyzed for the relation between contaminant fluxes at the outlet and mass removal. For the first time, this approach clearly allowed the differentiation of the treatments. As a conclusion, experiments showed that the most important differences between the tested treatment techniques were not the global mass removal rates but the time required to reach 99% decrease in

  20. Removal of contaminated asphalt layers by using heat generating powder metallic systems

    International Nuclear Information System (INIS)

    Barinov, A.S.; Karlina, O.K.; Ojovan, M.I.

    1996-01-01

    Heat generating systems on the base of powder metallic fuel were used for the removal of contaminated asphalt layers. Decontamination of spots which had complex geometric form was performed. Asphalt layers with deep contamination were removed essentially all radionuclides being retained in asphalt residue. Only a small part (1 - 2 %) of radionuclides could pass to combustion slag. No radionuclides were detected in aerosol-gas phase during decontamination process

  1. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco

    2016-01-01

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  2. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  3. Application of the Guided Wave Technique to the Heat Exchanger Tube in NPP

    International Nuclear Information System (INIS)

    Yang, Dong Soon; Kim, Hyung Nam; Yoo, Hyun Joo

    2005-01-01

    The heat exchanger tube is examined by the method of eddy current test(ECT) to identify the integrity of the nuclear power plant. Because ECT probe is moved through the tube inside to identify flaws, the ECT probe should be exchanged periodically due to the wear of probe surface in order to remove the noise form the ECT signal. Moreover, it is impossible to examine the tube by ECT method because the ECT probe can not move through the inside due to the deformation such as dent. Recently, the theory of guided wave was established and the equipment applying the theory has been actively developed so as to overcome the limitation of ECT method for the tube inspection of heater exchanger in nuclear power plant. The object of this study is to know the application of the guided wave technique to heat exchanger tube in NPP

  4. EFFECT OF HEAT-DISPERSING ON STICKIES AND THEIR REMOVAL IN POST-FLOTATION

    OpenAIRE

    Yang Gao,; Menghua Qin,; Hailong Yu,; Fengshan Zhang

    2012-01-01

    The effect of heat-dispersing on sticky substances in a deinking pulping line was studied under different conditions including varying temperature, disc clearance, and pulp consistency. Sticky substances were quantitatively investigated before and after the heat-dispersing, and categorized into macro-, mini-, and micro-stickies as well as dissolved and colloidal substances. Meanwhile, their extents of removal in post-flotation were evaluated. The results showed that raising temperature, reduc...

  5. Literature survey of heat transfer enhancement techniques in refrigeration applications

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, M.K.; Shome, B. [Rensselaer Polytechnic Inst., Troy, NY (United States). Dept. of Mechanical Engineering, Aeronautical Engineering and Mechanics

    1994-05-01

    A survey has been performed of the technical and patent literature on enhanced heat transfer of refrigerants in pool boiling, forced convection evaporation, and condensation. Extensive bibliographies of the technical literature and patents are given. Many passive and active techniques were examined for pure refrigerants, refrigerant-oil mixtures, and refrigerant mixtures. The citations were categorized according to enhancement technique, heat transfer mode, and tube or shell side focus. The effects of the enhancement techniques relative to smooth and/or pure refrigerants were illustrated through the discussion of selected papers. Patented enhancement techniques also are discussed. Enhanced heat transfer has demonstrated significant improvements in performance in many refrigerant applications. However, refrigerant mixtures and refrigerant-oil mixtures have not been studied extensively; no research has been performed with enhanced refrigerant mixtures with oil. Most studies have been of the parametric type; there has been inadequate examination of the fundamental processes governing enhanced refrigerant heat transfer, but some modeling is being done and correlations developed. It is clear that an enhancement technique must be optimized for the refrigerant and operating condition. Fundamental processes governing the heat transfer must be examined if models for enhancement techniques are to be developed; these models could provide the method to optimize a surface. Refrigerant mixtures, with and without oil present, must be studied with enhancement devices; there is too little known to be able to estimate the effects of mixtures (particularly NARMs) with enhanced heat transfer. Other conclusions and recommendations are offered.

  6. Heat removing device for nuclear reactor container facility

    Energy Technology Data Exchange (ETDEWEB)

    Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

    1994-09-30

    A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

  7. A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

    International Nuclear Information System (INIS)

    Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

    1995-01-01

    The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

  8. Meeting of Specialists on the Reliability of Decay Heat Removal Systems for Fast Reactors. Summary Report

    International Nuclear Information System (INIS)

    1975-10-01

    The Specialists Meeting on Reliability of Decay Heat Removal Systems proposed for Fast Reactors was sponsored by the UKAEA Safety & Reliability Directorate and held at Harwell between 28th April and 1st May, 1975. The meeting was attended by delegates from six countries - (USA, Federal Republic of Germany, France, Japan, USSR and the UK). A list of participants is included in an Appendix to this report. The subject matter of the meeting was concerned with the degree to which the ability to maintain decay heat removal from a fast reactor after shutdown in normal and abnormal circumstances could be guaranteed by design provisions and substantiated by reliability analysis techniques, operational testing etc. Consideration of conditions prevailing after a hypothetical core melt down incident were not included in the subject matter. The deliberations of the meeting were focussed at each working session on a defined theme and its dependant topics as shown in the detailed Agenda included in this report. Although provision had been made in the Agenda for a limited amount of discussion of the decay heat rejection problems of Gas Cooled Fast Reactors, delegates had no contributions to offer on this subject. During each session a Recording Secretary prepared a summary of the main points made by national delegates and of the resulting recommendations and conclusions. These draft summaries were made available to delegates during subsequent sessions of the meeting and approved by them for inclusion in the Summary, General Conclusions and Recommendations provided under Table of Contents (item 3 and 4)

  9. Heat Removal from Bipolar Transistor by Loop Heat Pipe with Nickel and Copper Porous Structures

    Directory of Open Access Journals (Sweden)

    Patrik Nemec

    2014-01-01

    Full Text Available Loop heat pipes (LHPs are used in many branches of industry, mainly for cooling of electrical elements and systems. The loop heat pipe is a vapour-liquid phase-change device that transfers heat from evaporator to condenser. One of the most important parts of the LHP is the porous wick structure. The wick structure provides capillary force to circulate the working fluid. To achieve good thermal performance of LHP, capillary wicks with high permeability and porosity and fine pore radius are expected. The aim of this work was to develop porous structures from copper and nickel powder with different grain sizes. For experiment copper powder with grain size of 50 and 100 μm and nickel powder with grain size of 10 and 25 μm were used. Analysis of these porous structures and LHP design are described in the paper. And the measurements’ influences of porous structures in LHP on heat removal from the insulated gate bipolar transistor (IGBT have been made.

  10. Tritium Removal from Carbon Plasma Facing Components

    International Nuclear Information System (INIS)

    Skinner, C.H.; Coad, J.P.; Federici, G.

    2003-01-01

    Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating

  11. Improved Design Concept for ensuring the Passive Decay Heat Removal Performance of an SFR

    International Nuclear Information System (INIS)

    Eoh, Jae Hyuk; Lee, Tae Ho; Han, Ji Woong; Kim, Seong O

    2011-01-01

    In order to enhance the operational reliability of a purely passive decay heat removal system in KALIMER, which is named as PDRC, three design options to prevent a sodium freezing in an intermediate decay heat removal circuit were proposed, and their feasibilities was quantitatively evaluated. For all the options, more specific design considerations were made to confirm their feasibility to properly materialize their concepts in a practical system design procedure, and the general definitions for a purely passive concept and its design features have been discussed. A numerical study to evaluate the coastdown flow effect of the primary pump was performed to figure out the early stage DHR capability inside reactor pool during a loss of normal heat sink accident. The thermal-hydraulic calculations have been made by using the COMMIX-1AR/P code, and it was found that the initiation of heat removal by DHX could be accelerated by the increase of the coastdown time but it needs a large-sized flywheel. For the demonstration of the innovative concept, a large scale sodium thermal-hydraulic test facility is currently being designed. It is very difficult to reproduce both a hydrodynamic and a thermodynamic similarity to the prototype plant if the thermal driving head is determined by structure-to-fluid heat transfer under natural circulation flow. Hence the similitude requirements for the sodium thermal-hydraulic test facility employing natural convection heat transfer were developed, and the preliminary design data of the test facility by implementing proper scaling methodologies was produced. The design restrictions imposed on the test facility and the scaling distortions of the design data to the full-scale system were also discussed

  12. Concept Design of a Gravity Core Cooling Tank as a Passive Residual Heat Removal System for a Research Reactor

    International Nuclear Information System (INIS)

    Lee, Kwonyeong; Chi, Daeyoung; Kim, Seong Hoon; Seo, Kyoungwoo; Yoon, Juhyeon

    2014-01-01

    A core downward flow is considered to use a plate type fuel because it is benefit to install the fuel in the core. If a flow inversion from a downward to upward flow in the core by a natural circulation is introduced within a high heat flux region of residual heat, the fuel fails instantly due to zero flow. Therefore, the core downward flow should be sufficiently maintained until the residual heat is in a low heat flux region. In a small power research reactor, inertia generated by a flywheel of the PCP can maintain a downward flow shortly and resolve the problem of a flow inversion. However, a high power research reactor more than 10 MW should have an additional method to have a longer downward flow until a low heat flux. Usually, other research reactors have selected an active residual heat removal system as a safety class. But, an active safety system is difficult to design and expensive to construct. A Gravity Core Cooling Tank (GCCT) beside the reactor pool with a Residual Heat Removal Pipe connecting two pools was developed and designed preliminarily as a passive residual heat removal system for an open-pool type research reactor. It is very simple to design and cheap to construct. Additionally, a non-safety, but active residual heat removal system is applied with the GCCT. It is a Pool Water Cooling and Purification System. It can improve the usability of the research reactor by removing the thermal waves, and purify the reactor pool, the Primary Cooling System, and the GCCT. Moreover, it can reduce the pool top radiation level

  13. Heat transport and afterheat removal for gas cooled reactors under accident conditions

    International Nuclear Information System (INIS)

    2001-01-01

    The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

  14. Nitrogen removal from digested slurries using a simplified ammonia stripping technique.

    Science.gov (United States)

    Provolo, Giorgio; Perazzolo, Francesca; Mattachini, Gabriele; Finzi, Alberto; Naldi, Ezio; Riva, Elisabetta

    2017-11-01

    This study assessed a novel technique for removing nitrogen from digested organic waste based on a slow release of ammonia that was promoted by continuous mixing of the digestate and delivering a continuous air stream across the surface of the liquid. Three 10-day experiments were conducted using two 50-L reactors. In the first two, nitrogen removal efficiencies were evaluated from identical digestates maintained at different temperatures (30°C and 40°C). At the start of the first experiment, the digestates were adjusted to pH 9 using sodium hydroxide, while in the second experiment pH was not adjusted. The highest ammonia removal efficiency (87%) was obtained at 40°C with pH adjustment. However at 40°C without pH adjustment, removal efficiencies of 69% for ammonia and 47% for total nitrogen were obtained. In the third experiment two different digestates were tested at 50°C without pH adjustment. Although the initial chemical characteristics of the digestates were different in this experiment, the ammonia removal efficiencies were very similar (approximately 85%). Despite ammonia removal, the pH increased in all experiments, most likely due to carbon dioxide stripping that was promoted by temperature and mixing. The technique proved to be suitable for removing nitrogen following anaerobic digestion of livestock manure because effective removal was obtained at natural pH (≈8) and 40°C, common operating conditions at typical biogas plants that process manure. Furthermore, the electrical energy requirement to operate the process is limited (estimated to be 3.8kWhm -3 digestate). Further improvements may increase the efficiency and reduce the processing time of this treatment technique. Even without these advances slow-rate air stripping of ammonia is a viable option for reducing the environmental impact associated with animal manure management. Copyright © 2017 Elsevier Ltd. All rights reserved.

  15. Experimental research on passive residual heat remove system for advanced PWR

    International Nuclear Information System (INIS)

    Huang Yanping; Zhuo Wenbin; Yang Zumao; Xiao Zejun; Chen Bingde

    2003-01-01

    The experimental and qualified results of MISAP in the research of passive residual heat remove system of advanced PWR performed in the Bubble physics and natural circulation laboratory in Nuclear Power Institute of China in the past ten years is overviewed. Further researches for engineering research and design are also suggested

  16. Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

    2015-09-15

    Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

  17. Heat transfer enhancement in cross-flow heat exchanger using vortex generator

    International Nuclear Information System (INIS)

    Yoo, S. Y.; Kwon, H. K.; Kim, B. C.; Park, D. S.; Lee, S. S.

    2003-01-01

    Fouling is very serious problem in heat exchanger because it rapidly deteriorates the performance of heat exchanger. Cross-flow heat exchanger with vortex generators is developed, which enhance heat transfer and reduce fouling. In the present heat exchanger, shell and baffle are removed from the conventional shell-and-tube heat exchanger. The naphthalene sublimation technique is employed to measure the local heat transfer coefficients. The experiments are performed for single circular tube, staggered array tube bank and in-line array tube bank with and without vortex generators. Local and average Nusselt numbers of single tube and tube bank with vortex generator are investigated and compared to those of without vortex generator

  18. Design of DC Conduction Pump for PGSFR Active Decay Heat Removal System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dehee; Hong, Jonggan; Lee, Taeho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    A DC conduction pump has been designed for the ADHRS of PGSFR. A VBA code developed by ANL was utilized to design and optimize the pump. The pump geometry dependent parameters were optimized to minimize the total current while meeting the design requirements. A double-C type dipole was employed to produce the calculated magnetic strength. Numerical simulations for the magnetic field strength and its distribution around the dipole and for the turbulent flow under magnetic force will be carried out. A Direct Current (DC) conduction Electromagnetic Pump (EMP) has been designed for Active Decay Heat Removal System (ADHRS) of PGSFR. The PGSFR has active as well as passive systems for the DHRS. The passive DHRS (PDHRS) works by natural circulation head and the ADHRS is driven by an EMP for the DHRS sodium loop and a blower for the finned-tube sodium-to-air heat exchanger (FHX). An Annular Linear Induction Pump (ALIP) can be also considered for the ADHRS, but DC conduction pump has been chosen. Selection basis of DHRS EMP is addressed and EMP design for single ADHRS loop with 1MWt heat removal capacity is introduced.

  19. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    International Nuclear Information System (INIS)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

  20. Study of passive residual heat removal system of a modular small PWR reactor

    International Nuclear Information System (INIS)

    Araujo, Nathália N.; Su, Jian

    2017-01-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

  1. A technique to remove the tensile instability in weakly compressible SPH

    Science.gov (United States)

    Xu, Xiaoyang; Yu, Peng

    2018-01-01

    When smoothed particle hydrodynamics (SPH) is directly applied for the numerical simulations of transient viscoelastic free surface flows, a numerical problem called tensile instability arises. In this paper, we develop an optimized particle shifting technique to remove the tensile instability in SPH. The basic equations governing free surface flow of an Oldroyd-B fluid are considered, and approximated by an improved SPH scheme. This includes the implementations of the correction of kernel gradient and the introduction of Rusanov flux into the continuity equation. To verify the effectiveness of the optimized particle shifting technique in removing the tensile instability, the impacting drop, the injection molding of a C-shaped cavity, and the extrudate swell, are conducted. The numerical results obtained are compared with those simulated by other numerical methods. A comparison among different numerical techniques (e.g., the artificial stress) to remove the tensile instability is further performed. All numerical results agree well with the available data.

  2. ALPHA - The long-term passive decay heat removal and aerosol retention program

    International Nuclear Information System (INIS)

    Guentay, S.; Varadi, G.; Dreier, J.

    1996-01-01

    The Paul Scherrer Institute initiated the major new experimental and analytical program ALPHA in 1990. The program is aimed at understanding the long-term decay heat removal and aerosol questions for the next generation of Passive Light Water Reactors. The ALPHA project currently includes four major items: the large-scale, integral system behaviour test facility PANDA, which will be used to examine multidimensional effects of the SBWR decay heat removal system; an investigation of the thermal hydraulics of natural convection and mixing in pools and large volumes (LINX); a separate-effects study of aerosols transport and deposition in plenum and tubes (AIDA); while finally, data from the PANDA facility and supporting separate effects tests will be used to develop and qualify models and provide validation of relevant system codes. The paper briefly reviews the above four topics and current status of the experimental facilities. (author). 3 refs, 12 figs

  3. ALPHA - The long-term passive decay heat removal and aerosol retention program

    Energy Technology Data Exchange (ETDEWEB)

    Guentay, S; Varadi, G; Dreier, J [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-12-01

    The Paul Scherrer Institute initiated the major new experimental and analytical program ALPHA in 1990. The program is aimed at understanding the long-term decay heat removal and aerosol questions for the next generation of Passive Light Water Reactors. The ALPHA project currently includes four major items: the large-scale, integral system behaviour test facility PANDA, which will be used to examine multidimensional effects of the SBWR decay heat removal system; an investigation of the thermal hydraulics of natural convection and mixing in pools and large volumes (LINX); a separate-effects study of aerosols transport and deposition in plenum and tubes (AIDA); while finally, data from the PANDA facility and supporting separate effects tests will be used to develop and qualify models and provide validation of relevant system codes. The paper briefly reviews the above four topics and current status of the experimental facilities. (author). 3 refs, 12 figs.

  4. Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Lofaro, R.J.; Aggarwal, S.

    1992-01-01

    The effects of aging on Residual Heat Removal systems in Boiling Water Reactors have been studied as part of the Nuclear Plant Aging Research Program. The aging phenomena has been characterized by analyzing operating experience from various national data bases. In addition, actual plant data was obtained to supplement and validate the data base findings

  5. Effectiveness of photocatalytic filter for removing volatile organic compounds in the heating, ventilation, and air conditioning system.

    Science.gov (United States)

    Yu, Kuo-Pin; Lee, Grace Whei-May; Huang, Wei-Ming; Wu, Chih-Cheng; Lou, Chia-ling; Yang, Shinhao

    2006-05-01

    Nowadays, the heating, ventilation, and air conditioning (HVAC) system has been an important facility for maintaining indoor air quality. However, the primary function of typical HVAC systems is to control the temperature and humidity of the supply air. Most indoor air pollutants, such as volatile organic compounds (VOCs), cannot be removed by typical HVAC systems. Thus, some air handling units for removing VOCs should be added in typical HVAC systems. Among all of the air cleaning techniques used to remove indoor VOCs, photocatalytic oxidation is an attractive alternative technique for indoor air purification and deodorization. The objective of this research is to investigate the VOC removal efficiency of the photocatalytic filter in a HVAC system. Toluene and formaldehyde were chosen as the target pollutants. The experiments were conducted in a stainless steel chamber equipped with a simplified HVAC system. A mechanical filter coated with Degussa P25 titania photocatalyst and two commercial photocatalytic filters were used as the photocatalytic filters in this simplified HVAC system. The total air change rates were controlled at 0.5, 0.75, 1, 1.25, and 1.5 hr(-1), and the relative humidity (RH) was controlled at 30%, 50%, and 70%. The ultraviolet lamp used was a 4-W, ultraviolet-C (central wavelength at 254 nm) strip light bulb. The first-order decay constant of toluene and formaldehyde found in this study ranged from 0.381 to 1.01 hr(-1) under different total air change rates, from 0.34 to 0.433 hr(-1) under different RH, and from 0.381 to 0.433 hr(-1) for different photocatalytic filters.

  6. Experimental and analytical studies for the validation of HTR-VGD and primary cell passive decay heat removal. Supplement. Calculations

    International Nuclear Information System (INIS)

    Geiss, M.; Giannikos, A.; Hejzlar, P.; Kneer, A.

    1993-04-01

    The alternative concept for a modular HTR-reactor design by Siempelkamp, Krefeld, using a prestressed cast iron vessel (VGD) combined with a cast iron/concrete module for the primary cell with integrated passive decay heat removal system was fully qualified with respect to operational and accidental thermal loads. The main emphasis was to confirm and validate the passive decay heat removal capability. An experimental facility (INWA) was designed, instrumented and operated with an appropriate electrical heating system simulating steady-state operational and transient accidental thermal loads. The experiments were accompanied by extensive computations concerning the combination of conductive, radiative and convective energy transport mechanisms in the different components of the VGD/primary cell structures, as well as elastic-plastic stress analyses of the VGD. In addition, a spectrum of potential alternatives for passive energy removed options have been parametrically examined. The experimental data clearly demonstrate that the proposed Siempelkamp-design is able to passively and safely remove the decay heat for operational and accidental conditions without invalidating technological important thermal limits. This also holds in case of failures of both the natural convection system and ultimate heat sink by outside concrete water film cooling. (orig./HP) [de

  7. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

    For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

  8. Studies related to emergency decay heat removal in EBR-II

    International Nuclear Information System (INIS)

    Singer, R.M.; Gillette, J.L.; Mohr, D.; Tokar, J.V.; Sullivan, J.E.; Dean, E.M.

    1979-01-01

    Experimental and analytical studies related to emergency decay heat removal by natural circulation in the EBR-II heat transport circuits are described. Three general categories of natural circulation plant transients are discussed and the resultant reactor flow and temperature response to these events are presented. these categories include the following: (1) loss of forced flow from decay power and low initial flow rates; (2) reactor scram with a delayed loss of forced flow; and (3) loss of forced flow with a plant protective system activated scram. In all cases, the transition from forced to natural convective flow was smooth and the peak in-core temperature rises were small to moderate. Comparisons between experimental measurements in EBR-II and analytical predictions of the NATDEMO code are included

  9. Evaluation of the decay heat removal capability using the concept of a thermosyphon in the liquid metal reactor

    International Nuclear Information System (INIS)

    Kim, Y. S.; Sim, Y. S.; Kim, W. K.

    2000-01-01

    A study related to understand the characteristics of the heat pipe and thermosyphon was performed to evaluate their applicabilities to the current PSDRS (Passive Safety Decay heat Removal System) in the KALIMER (Korea Advanced LIquid MEtal Reactor) design. The possible heat transfer rate by the heat pipe and thermosyphon was reviewed to compare the required capability in the PSDRS. A quantitative comparison was done between the current PSDRS and the modified PSDRS with the thermosyphon. The result showed the dominant heat transfer rate in the air channel, e.g. radiation or convection, is different from each other. The total heat transfer rate is not sensitive to the operating temperature of the thermosyphon. The heat removal by the air in the modified case is relatively reduced and the resultant outlet temperature appears less than above 10 .deg. C. A reversal heat transfer between the air and the thermosyphon may exist near the exit of the active heat transfer region. The total heat transfer rate by the modified case showed about 20∼40% increase relative to the reference one

  10. Feasibility study for a postaccident heat removal facility

    International Nuclear Information System (INIS)

    Barts, E.W.; Apperson, C.E. Jr.; Dunwoody, W.E.; Bennett, J.G.

    1978-01-01

    An initial feasibility investigation for PAHRTEF, a Postaccident Heat Removal Test Facility, is presented. The facility would provide an experimental capability for PAHR experiments beyond that available in any currently existing or proposed U.S. safety test facility. The facility design presented in this report is based upon the technology developed for the ROVER nuclear rocket propulsion program. The core is a graphite-moderated, helium-cooled, epithermal core with radial reflector control. The PAHR experiments are located just below the reactor containment vessel, very near the bottom of the core. The experiments (up to 55% enriched) are driven and controlled by neutrons leaking axially from the core such that the PAHRTEF core and the experiment form a coupled reactor system. The experiment can be designed so that it is extremely unlikely that the test fuel by itself could form a critical system. The investigation indicates that adequate fission heating of large PAHR experiments could be provided at low driver core power levels. Both the reactor and the experiment handling and examination equipment can use available technology and, whenever possible, existing equipment and buildings

  11. Heat removal tests on dry storage facilities for nuclear spent fuels

    International Nuclear Information System (INIS)

    Wataru, M.; Saegusa, T.; Koga, T.; Sakamoto, K.; Hattori, Y.

    1999-01-01

    In Japan, spent fuel generated in NPP is controlled and stored in dry storage facility away-from reactor. Natural convection cooling system of the storage facility is considered advantageous from both safety and economic point of view. In order to realize this type of facility it is necessary to develop an evaluation method for natural convection characteristics and to make a rational design taking account safety and economic factors. Heat removal tests with the reduces scale models of storage facilities (cask, vault and silo) identified the the flow pattern in the test modules. The temperature and velocity distributions were obtained and the heat transfer characteristics were evaluated

  12. Design of SMART waste heat removal dry cooling tower using solar energy

    International Nuclear Information System (INIS)

    Choi, Yong Jae; Jeong, Yong Hoon

    2014-01-01

    The 85% of cooling system are once-through cooling system and closed cycle wet cooling system. However, many countries are trying to reduce the power plant water requirement due to the water shortage and water pollution. Dry cooling system is investigated for water saving advantage. There are two dry cooling system which are direct and indirect cooling system. In direct type, turbine exhaust is directly cooled by air-cooled condenser. In indirect system, turbine steam is cooled by recirculating intermediate cooling water loop, then the loop is cooled by air-cooled heat exchanger in cooling tower. In this paper, the purpose is to remove SMART waste heat, 200MW by using newly designed tower. The possibility of enhancing cooling performance by solar energy is analyzed. The simple cooling tower and solar energy cooling tower are presented and two design should meet the purpose of removing SMART waste heat, 200MW. In first design, when tower diameter is 70m, the height of tower should be 360m high. In second design, the chimney height decrease from 360m to 180m as collector radius increase from 100m to 500m due to collector temperature enhancement by solar energy, To analyze solar cooling tower further, consideration of solar energy performance at night should be analyzed

  13. Design of SMART waste heat removal dry cooling tower using solar energy

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Jae; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    The 85% of cooling system are once-through cooling system and closed cycle wet cooling system. However, many countries are trying to reduce the power plant water requirement due to the water shortage and water pollution. Dry cooling system is investigated for water saving advantage. There are two dry cooling system which are direct and indirect cooling system. In direct type, turbine exhaust is directly cooled by air-cooled condenser. In indirect system, turbine steam is cooled by recirculating intermediate cooling water loop, then the loop is cooled by air-cooled heat exchanger in cooling tower. In this paper, the purpose is to remove SMART waste heat, 200MW by using newly designed tower. The possibility of enhancing cooling performance by solar energy is analyzed. The simple cooling tower and solar energy cooling tower are presented and two design should meet the purpose of removing SMART waste heat, 200MW. In first design, when tower diameter is 70m, the height of tower should be 360m high. In second design, the chimney height decrease from 360m to 180m as collector radius increase from 100m to 500m due to collector temperature enhancement by solar energy, To analyze solar cooling tower further, consideration of solar energy performance at night should be analyzed.

  14. A study of emergency American football helmet removal techniques.

    Science.gov (United States)

    Swartz, Erik E; Mihalik, Jason P; Decoster, Laura C; Hernandez, Adam E

    2012-09-01

    The purpose was to compare head kinematics between the Eject Helmet Removal System and manual football helmet removal. This quasi-experimental study was conducted in a controlled laboratory setting. Thirty-two certified athletic trainers (sex, 19 male and 13 female; age, 33 ± 10 years; height, 175 ± 12 cm; mass, 86 ± 20 kg) removed a football helmet from a healthy model under 2 conditions: manual helmet removal and Eject system helmet removal. A 6-camera motion capture system recorded 3-dimensional head position. Our outcome measures consisted of the average angular velocity and acceleration of the head in each movement plane (sagittal, frontal, and transverse), the resultant angular velocity and acceleration, and total motion. Paired-samples t tests compared each variable across the 2 techniques. Manual helmet removal elicited greater average angular velocity in the sagittal and transverse planes and greater resultant angular velocity compared with the Eject system. No differences were observed in average angular acceleration in any single plane of movement; however, the resultant angular acceleration was greater during manual helmet removal. The Eject Helmet Removal System induced greater total head motion. Although the Eject system created more motion at the head, removing a helmet manually resulted in more sudden perturbations as identified by resultant velocity and acceleration of the head. The implications of these findings relate to the care of all cervical spine-injured patients in emergency medical settings, particularly in scenarios where helmet removal is necessary. Copyright © 2012 Elsevier Inc. All rights reserved.

  15. Probabilistic analysis of the loss of the decay heat removal function for Creys-Malville reactor

    International Nuclear Information System (INIS)

    Lanore, J.M.; Villeroux-Lombard, C.; Bouscatie, F.; Pavret de la Rochefordiere, A.

    1982-01-01

    The classical fault tree/event tree methods do not take into account the dependence in time of the systems behaviour during the sequences, and that is quite unrealistic for the decay heat removal function. It was then necessary to use a new methodology based on functional states of the whole system and on transition laws between these states. Thus, the probabilistic analysis of the decay heat removal function for Creys-Malville plant is performed in a global way. The main accident sequences leading to the loss of the function are then determined a posteriori. The weak points are pointed out, in particular the importance of common mode failures

  16. Residual heat removal during accidental situations

    International Nuclear Information System (INIS)

    Depond, M.; Sureau, H.; Tellier, N.

    1983-07-01

    Existing emergency procedures, whose purpose is residual heat removal and a safe recovery are based on sequential analysis and initiating event diagnosis. This approach was found in some cases inappropriate and inefficient, specially in case of out-of-design accidents corresponding to multiple equipment failure or simultaneous human failures. To cope with these situations, a new approach was necessary. Parallel studies performed in France at Framatome (the designer) and Electricity de France (the utility) gave a new method, called NSSS physical states approach. Prior to the implementation of this method which necessitates further studies and developments, some improvements in the existing operating procedures derived from the NSSS physical states have already been achieved: that is the case for the safety injection control and the development of an emergency procedure called ''U1''. This paper will briefly physical states approach and present the ''U1'' procedure. The tools which will be used to chack these methods are also mentioned

  17. A decay heat removal system requiring no external energy

    International Nuclear Information System (INIS)

    Costes, D.; Fermandjian, J.

    1983-12-01

    A new Decay heat Removal System is described for PWR's with dry containment, i.e. a containment building which encloses no permanent reserve of cooling water. This new system is intended to provide a high level of safety since it uses no external energy, but only the thermodynamic energy of the air-steam-liquid water mixture generated in the containment after the failure of the primary circuit (''LOCA'') or of the secondary circuit. Thermodynamics of the system is evaluated first: after some design considerations, the use of the system for protecting actual PWR's is addressed

  18. Towards convective heat transfer enhancement: surface modification, characterization and measurement techniques

    NARCIS (Netherlands)

    Taha, T.J.; Thakur, D.B.; van der Meer, Theodorus H.

    2012-01-01

    In this work, heat transfer surface modification and heat transfer measurement technique is developed. Heat transfer investigation was aimed to study the effect of carbon nano fibers (extremely high thermal conductive material) on the enhancement level in heat transfer. Synthesis of these carbon

  19. Techniques for the improvement of heat exchange

    International Nuclear Information System (INIS)

    Huyghe, J.

    1983-01-01

    The target of the described techniques is either to improve the performances of the equipements or to reduce their cost price. These techniques were developed at the laboratory scale and some of them were tested in pilot units. Neither of them has yet been applied in commercial plants. Different type of corrogated tubes, thin film evaporation, dropwise condensation, plastic tube or plates, tube configuration and heat pipes are examined [fr

  20. Pectus bar removal: surgical technique and strategy to avoid complications.

    Science.gov (United States)

    Park, Hyung Joo; Kim, Kyung Soo

    2016-01-01

    Pectus bar removal is the final stage of the procedure for minimally invasive repair of pectus excavatum. Based on our experience with one of the largest scale data, we would like to address the important issues in pectus bar removal, such as appropriate duration of bar maintenance, techniques for bar removal, and strategies to avoid complications. Between September 1999 and August 2015, we operated on 2,553 patients with pectus excavatum and carinatum using pectus bars for a minimally invasive approach. Among them, 1,821 patients (71.3%) underwent pectus bar removal as a final stage of pectus deformity repair, and their data were analyzed retrospectively to identify the outcomes and adverse effects of the pectus bar removal procedure. The mean age of the patients was 9.13 years (range, 16 months to 44 years) and the male to female ratio was 3.55. The study is approved by the Institutional Review Board (IRB), the ethical committee of Seoul St. Mary's Hospital. The IRB has exempted the informed consent from every patient in this study due to this is a retrospective chart review without revealing any patients' personal data. Our technique involved straightening of the bar in a supine position. The overall mean duration of pectus bar maintenance was 2.57 years (range, 4 months to 14 years). The mean duration was 2.02 years (range, 4 months to 7 years) for children under 12 years, 2.99 years (range, 7 months to 9 years) for teenagers aged 12-20 years, and 3.53 years (range, 3 months to 14 years) for adults over 20 years. Forty-eight patients (2.6%) underwent bar removal more than 5 years after bar insertion and 58 patients (3.2%) underwent bar removal earlier than initially planned. The most common adverse reaction after bar removal was wound seroma including infection (43 patients, 2.36%). Recurrence after bar removal occurred in nine patients (0.49%), and seven of these required redo repair (0.38%). Pectus bar removal is a safe and straightforward procedure with a

  1. Experimental study on heat transfer augmentation for high heat flux removal in rib-roughened narrow channels

    Energy Technology Data Exchange (ETDEWEB)

    Islam, M.S.; Monde, Masanori [Saga Univ. (Japan); Hino, Ryutaro; Haga, Katsuhiro; Sudo, Yukio

    1997-07-01

    Frictional pressure drop and heat transfer performance in a very narrow rectangular channel having one-sided constant heat flux and repeated-ribs for turbulent flow have been investigated experimentally, and their experimental correlations were obtained using the least square method. The rib pitch-to-height ratios(p/k) were 10 and 20 while holding the rib height constant at 0.2mm, the Reynolds number(Re) from 2,414 to 98,458 under different channel heights of 1.2mm, 2.97mm, and 3.24mm, the rib height-to-channel equivalent diameter(k/De) of 0.03, 0.04, and 0.09 respectively. The results show that the rib-roughened surface augments heat transfer 2-3 times higher than that of the smooth surface with the expense of 2.8-4 times higher frictional pressure drop under Re=5000-10{sup 5}, p/k=10, and H=1.2mm. Experimental results obtained by channel height, H=1.2mm shows a little bit higher heat transfer and friction factor performance than the higher channel height, H=3.24mm. The effect of fin and consequently higher turbulence intensity are responsible for producing higher heat transfer rates. The obtained correlations could be used to design the cooling passages between the target plates to remove high heat flux up to 12MW/m{sup 2} generated at target plates in a high-intensity proton accelerator system. (author). 54 refs.

  2. Experimental study on heat transfer augmentation for high heat flux removal in rib-roughened narrow channels

    International Nuclear Information System (INIS)

    Islam, M.S.; Monde, Masanori; Hino, Ryutaro; Haga, Katsuhiro; Sudo, Yukio.

    1997-07-01

    Frictional pressure drop and heat transfer performance in a very narrow rectangular channel having one-sided constant heat flux and repeated-ribs for turbulent flow have been investigated experimentally, and their experimental correlations were obtained using the least square method. The rib pitch-to-height ratios(p/k) were 10 and 20 while holding the rib height constant at 0.2mm, the Reynolds number(Re) from 2,414 to 98,458 under different channel heights of 1.2mm, 2.97mm, and 3.24mm, the rib height-to-channel equivalent diameter(k/De) of 0.03, 0.04, and 0.09 respectively. The results show that the rib-roughened surface augments heat transfer 2-3 times higher than that of the smooth surface with the expense of 2.8-4 times higher frictional pressure drop under Re=5000-10 5 , p/k=10, and H=1.2mm. Experimental results obtained by channel height, H=1.2mm shows a little bit higher heat transfer and friction factor performance than the higher channel height, H=3.24mm. The effect of fin and consequently higher turbulence intensity are responsible for producing higher heat transfer rates. The obtained correlations could be used to design the cooling passages between the target plates to remove high heat flux up to 12MW/m 2 generated at target plates in a high-intensity proton accelerator system. (author). 54 refs

  3. A modified-simple technique of removing the lens cortex during cataract surgery

    Directory of Open Access Journals (Sweden)

    Kyung Eun Han

    2017-01-01

    Full Text Available We describe here a surgical technique of removing the remaining cortex after phacoemulsification without performing the conventional irrigation/aspiration (I/A procedure. In this technique, the remaining cortex attached to the posterior capsule was separated and dissected into several pieces by continuous irrigation with balanced salt solution, which was supplied through a syringe attached to a bent, blunt-tip needle. Approximately, 10 s of manual irrigation separated most of the remaining cortex from the posterior capsule. Then, the capsular bag was inflated with an ophthalmic viscoelastic device (OVD, and this pushed the separated cortex toward the capsular fornix mechanically. An intraocular lens was inserted into the capsular bag, following which the remaining cortex and OVD were removed concomitantly using an automated I/A handpiece. This technique is a simple and easy maneuver to remove the cortex from all areas, including the subincisional area, and reduce the possibility of a posterior capsule tear.

  4. Reliability assessment on decay heat removal system of a fast reactor

    International Nuclear Information System (INIS)

    Hioki, Kazumasa

    1991-01-01

    The reliability of a decay heat removal system (DHRS) is influenced by the success criteria, the components which constitute the system, the support systems configuration, and the mission time. Assessments were performed to investigate quantitatively the effects of these items. Failure probabilities of DHRS under forced or natural circulation modes were calculated and then components and systems of large importance for each mode were identified. (author)

  5. Steam generator concept of a small HTR for reheating and for removal of the residual heat

    Energy Technology Data Exchange (ETDEWEB)

    Singh, J; Barnert, H; Hohn, H; Mondry, M [Institut fuer Reaktorenentwicklung, Kernforschungsanlage Juelich GmbH, Juelich (Germany)

    1988-07-01

    The steam generator of a small HTR is arranged above the core in an in line design of the primary loop, thereby helium flows upwards. Water flows downwards in the steam generator to realize cross flow. To achieve stable evaporation conditions during part load operation it is desired to realize upward evaporation in the steam generator. Moreover if the steam generator is also used as a heat sink for removal of residual heat, this desire of upwards evaporation becomes more imperative. It is possible to realize the design of steam generator with upwards evaporation by arranging a hot gas duct in its central region, so that hot helium can flow upwards through it. Therefore helium enters the steam generator from the top and flows downwards and water upwards. In the presented design, a heat exchanger is arranged in the central region of the steam generator instead of a hot gas duct. Hot helium of 750 deg. C flows upwards in this heat exchanger and thereby cools down to the temperature of about 700 deg. C before it enters the bundle of the steam generator at the top. Through an intermediate loop this heat is transferred outside the primary loop, where in an extra heat exchanger live steam is reheated to improve the thermal efficiency of the plant. This intermediate loop works on the basis of forced convection and transfer about 25 MW for reheating. During the shutdown operation of the reactor, this heat exchanger in the central region of the steam generator serves as a heat sink for removal of the residual heat through natural convection in the primary loop. At the same time it is further possible, that intermediate loop also works on the basis of natural convection, because during shutdown operation only a very small amount of heat has to be removed and moreover the outside heat exchanger can be arranged much higher above the central heat exchanger to get favourable conditions for the natural convection. Some of the highlights of the central heat exchanger are: coaxial

  6. Heat removal capability of divertor coaxial tube assembly

    International Nuclear Information System (INIS)

    Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

    1994-05-01

    To deal with high power flowing in the divertor region, an advanced divertor concept with gas target has been proposed for use in ITER/EDA. The concept uses a divertor channel to remove the radiated power while allowing neutrals to recirculate. Candidate channel wall designs include a tube array design where many coaxial tubes are arranged in the toroidal direction to make louver. The coaxial tube consists of a Be protection tube encases many supply tubes wound helically around a return tube. V-alloy and hardened Cu-alloy have been proposed for use in the supply and return tubes. Some coolants have also been proposed for the design including pressurized He and liquid metals, because these coolants are consistent with the selection of coolants for the blanket and also meet the requirement of high temperature operation. In the coaxial tube design, the coolant area is restricted and brittle Be material is used under severe thermal cyclings. Thus, to obtain the coaxial tube with sufficient safety margin for the expected fusion power excursion, it is essential to understand its applicability limit. The paper discusses heat removal capability of the coaxial tube and recommends some design modifications. (author)

  7. VOC removal by microwave, electron beam and catalyst technique

    International Nuclear Information System (INIS)

    IghigeanuI, D.; Martin, D.; OproiuI, C.; Manaila, E.; Craciun, G.; Calinescu, I.; Zissulescu, E.

    2007-01-01

    A hybrid technique, developed for VOCs removal using microwave (MW) treatment, electron beam (EB) irradiation and catalyst method, is presented. Two hybrid laboratory installations, developed for the study of air pollution control by combined EB irradiation, MW irradiation and catalyst, are described. Air loaded with toluene was treated at different MW power levels, water content, flow rates, and different irradiation modes, separately and combined with MW and EB. Also, simultaneous EB and MW irradiation method was applied to SO 2 and NO x removal. Real synergy effects between EB induced NTP, MW induced NTP and catalysis can be observed

  8. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Alvim, Antonio C.M.

    2017-01-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  9. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail: thiagodbtr@gmail.com, E-mail: lapa@ien.gov.br, E-mail: alvim@nuclear.ufrj.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  10. Computed tomography assessment of the efficiency of different techniques for removal of root canal filling material

    International Nuclear Information System (INIS)

    Dall'agnol, Cristina; Barletta, Fernando Branco; Hartmann, Mateus Silveira Martins

    2008-01-01

    This study evaluated the efficiency of different techniques for removal of filling material from root canals, using computed tomography (CT). Sixty mesial roots from extracted human mandibular molars were used. Root canals were filled and, after 6 months, the teeth were randomly assigned to 3 groups, according to the root-filling removal technique: Group A - hand instrumentation with K-type files; Group B - reciprocating instrumentation with engine-driven K-type files; and Group C rotary instrumentation with engine-driven ProTaper system. CT scans were used to assess the volume of filling material inside the root canals before and after the removal procedure. In both moments, the area of filling material was outlined by an experienced radiologist and the volume of filling material was automatically calculated by the CT software program. Based on the volume of initial and residual filling material of each specimen, the percentage of filling material removed from the root canals by the different techniques was calculated. Data were analyzed statistically by ANOVA and chi-square test for linear trend (α=0.05). No statistically significant difference (p=0.36) was found among the groups regarding the percent means of removed filling material. The analysis of the association between the percentage of filling material removal (high or low) and the proposed techniques by chi-square test showed statistically significant difference (p=0.015), as most cases in group B (reciprocating technique) presented less than 50% of filling material removed (low percent removal). In conclusion, none of the techniques evaluated in this study was effective in providing complete removal of filling material from the root canals. (author)

  11. Computed tomography assessment of the efficiency of different techniques for removal of root canal filling material

    Energy Technology Data Exchange (ETDEWEB)

    Dall' agnol, Cristina; Barletta, Fernando Branco [Lutheran University of Brazil, Canoas, RS (Brazil). Dental School. Dept. of Dentistry and Endodontics]. E-mail: fbarletta@terra.com.br; Hartmann, Mateus Silveira Martins [Uninga Dental School, Passo Fundo, RS (Brazil). Postgraduate Program in Dentistry

    2008-07-01

    This study evaluated the efficiency of different techniques for removal of filling material from root canals, using computed tomography (CT). Sixty mesial roots from extracted human mandibular molars were used. Root canals were filled and, after 6 months, the teeth were randomly assigned to 3 groups, according to the root-filling removal technique: Group A - hand instrumentation with K-type files; Group B - reciprocating instrumentation with engine-driven K-type files; and Group C rotary instrumentation with engine-driven ProTaper system. CT scans were used to assess the volume of filling material inside the root canals before and after the removal procedure. In both moments, the area of filling material was outlined by an experienced radiologist and the volume of filling material was automatically calculated by the CT software program. Based on the volume of initial and residual filling material of each specimen, the percentage of filling material removed from the root canals by the different techniques was calculated. Data were analyzed statistically by ANOVA and chi-square test for linear trend ({alpha}=0.05). No statistically significant difference (p=0.36) was found among the groups regarding the percent means of removed filling material. The analysis of the association between the percentage of filling material removal (high or low) and the proposed techniques by chi-square test showed statistically significant difference (p=0.015), as most cases in group B (reciprocating technique) presented less than 50% of filling material removed (low percent removal). In conclusion, none of the techniques evaluated in this study was effective in providing complete removal of filling material from the root canals. (author)

  12. One-dimensional nonlinear inverse heat conduction technique

    International Nuclear Information System (INIS)

    Hills, R.G.; Hensel, E.C. Jr.

    1986-01-01

    The one-dimensional nonlinear problem of heat conduction is considered. A noniterative space-marching finite-difference algorithm is developed to estimate the surface temperature and heat flux from temperature measurements at subsurface locations. The trade-off between resolution and variance of the estimates of the surface conditions is discussed quantitatively. The inverse algorithm is stabilized through the use of digital filters applied recursively. The effect of the filters on the resolution and variance of the surface estimates is quantified. Results are presented which indicate that the technique is capable of handling noisy measurement data

  13. Heat removal tests for pressurized water reactor containment spray by largescale facility

    International Nuclear Information System (INIS)

    Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

    1983-01-01

    Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

  14. Possible design of PBR for passive decay heat removal

    International Nuclear Information System (INIS)

    Sambuu, Odmaa; Obara, Toru

    2016-01-01

    Conditions for design parameters of above-ground and underground, prismatic high-temperature gas-cooled reactor (HTGR)s for passive decay heat removal based on fundamental heat transfer mechanisms were obtained in the previous works. In the present study, analogous conditions were obtained for pebble bed reactors by performing the same procedure using the model for heat transfer in porous media of COMSOL 4.3a software, and the results were compared. For the power density profile, several approximated distributions together with original one throughout the 10-MWt high-temperature gas-cooled reactor-test module (HTR-10) were used, and it was found that an HTR-10 with a uniform power density profile has the higher safety margin than those with other profiles. In other words, the safety features of a PBR can be enhanced by flattening the power density profile. We also found that a prismatic HTGR with a uniform power density profile throughout the core has a greater safety margin than a PBR with the same design characteristics. However, when the power density profile is not flattened during the operation, the PBR with the linear power density profile has more safety margin than the prismatic HTGR with the same design parameters and with the power density profile by cosine and Bessel functions. (author)

  15. Techniques for measurement of heat flux in furnace waterwalls of boilers and prediction of heat flux – A review

    International Nuclear Information System (INIS)

    Sankar, G.; Chandrasekhara Rao, A.; Seshadri, P.S.; Balasubramanian, K.R.

    2016-01-01

    Highlights: • Heat flux measurement techniques applicable to boiler water wall are elaborated. • Applications involving heat flux measurement in boiler water wall are discussed. • Appropriate technique for usage in high ash Indian coal fired boilers is required. • Usage of chordal thermocouple is suggested for large scale heat flux measurements. - Abstract: Computation of metal temperatures in a furnace waterwall of a boiler is necessary for the proper selection of tube material and thickness. An adequate knowledge of the heat flux distribution in the furnace walls is a prerequisite for the computation of metal temperatures. Hence, the measurement of heat flux in a boiler waterwall is necessary to arrive at an optimum furnace design, especially for high ash Indian coal fired boilers. Also, a thoroughly validated furnace model will result in a considerable reduction of the quantum of experimentation to be carried out. In view of the above mentioned scenario, this paper reviews the research work carried out by various researchers by experimentation and numerical simulation in the below mentioned areas: (i) furnace modeling and heat flux prediction, (ii) heat flux measurement techniques and (iii) applications of heat flux measurements.

  16. Removal of Polypropylene Sling Mesh From the Urethra: An Anatomic Technique.

    Science.gov (United States)

    Freilich, Drew A; Rovner, Eric S

    2015-07-01

    To describe a technique for removal of intraurethral mesh with minimal disruption of the periurethral anatomy. Through a midline transvaginal approach the sling is located lateral to the urethra and divided. The medial portion of the divided sling is carefully dissected back to its entrance laterally into the urethral lumen. A stay suture is placed on the dissected sling. The sling is located on the contralateral side and likewise divided and dissected back to the urethral lumen. The completely dissected sling is pulled through such that the holding stitch is through and through the urethral lumen, allowing easy localization of the urethral defect on lateral walls of the urethra. These defects are closed with an absorbable suture and the vaginal incision is closed. Three patients have undergone a transvaginal removal of their intraurethral mesh using the described technique. At a mean follow-up of 6.0 months, there were no intraoperative or postoperative complications. All patients were obstructed preoperatively and all developed stress urinary incontinence postoperatively requiring 0-1 pads per day. Current approaches to the surgical repair of chronic intraurethral mesh have significant limitations that are minimized by the described technique. This anatomic removal of mesh from the urethra has several advantages including no disruption of the ventral wall of the urethra, complete removal of foreign body from the urethra, and simplified localization of the urethral wall defect to allow for anatomic closure. Copyright © 2015 Elsevier Inc. All rights reserved.

  17. Design and modeling of an advanced marine machinery system including waste heat recovery and removal of sulphur oxides

    DEFF Research Database (Denmark)

    Frimann Nielsen, Rasmus; Haglind, Fredrik; Larsen, Ulrik

    2014-01-01

    the efficiency of machinery systems. The wet sulphuric acid process is an effective way of removing flue gas sulphur oxides from land-based coal-fired power plants. Moreover, organic Rankine cycles (ORC) are suitable for heat to power conversion for low temperature heat sources. This paper describes the design...... that an ORC placed after the conventional waste heat recovery system is able to extract the sulphuric acid from the exhaust gas, while at the same time increase the combined cycle thermal efficiency by 2.6%. The findings indicate that the technology has potential in marine applications regarding both energy...... and modeling of a highly efficient machinery system which includes the removal of exhaust gas sulphur oxides. The system consists of a two-stroke diesel engine, the wet sulphuric process for sulphur removal, a conventional steam Rankine cycle and an ORC. Results of numerical modeling efforts suggest...

  18. 'Stent in a stent'--an alternative technique for removing partially covered stents following sleeve gastrectomy complications.

    Science.gov (United States)

    Vasilikostas, Georgios; Sanmugalingam, Nimalan; Khan, Omar; Reddy, Marcus; Groves, Chris; Wan, Andrew

    2014-03-01

    Endoscopic stenting is a relatively new technique for the treatment of post sleeve gastrectomy complications. Partially covered stents are used in this method to minimise the risk of migration but they are associated with difficulties with removal. Patients requiring emergency stenting following sleeve gastrectomy underwent insertion of a partially covered metallic stent. One month later, if the stent was not easily removable, a fully covered overlapping stent was inserted and the patient was readmitted 2 weeks later for removal of both stents. Four patients required stenting following sleeve gastrectomy leaks, and one patient required stenting for a stricture. In these cases, a 'stent in a stent' technique was used for removal. This technique allows the safe removal of partially covered stents inserted following sleeve gastrectomy complications.

  19. A Novel Cast Removal Training Simulation to Improve Patient Safety.

    Science.gov (United States)

    Brubacher, Jacob W; Karg, Jeffrey; Weinstock, Peter; Bae, Donald S

    2016-01-01

    Cast application and removal are essential to orthopedics and performed by providers of variable training. Simulation training and practice of proper cast application and removal may reduce injury, optimize outcomes, and reduce health care costs. The purpose of this educational initiative was to develop, validate, and implement a novel simulation trainer and curriculum to improve safety during cast removal. In all, 30 thermocouples (Omega, Stamford, CT) were applied to a radius fracture model (Sawbones, Vashon, WA). After reduction and cast application, a saw (Stryker, Kalamazoo, MI) was used to cut the cast with temperature recording. Both "good" and "poor" techniques-as established by consensus best practices-were used. Maximal temperatures were compared to known thresholds for thermal injury; humans experience pain at temperatures exceeding 47°C and contact temperatures exceeding 60°C may lead to epidermal necrosis. Construct validity was evaluated by assessing novice (postgraduate year 1), intermediate (postgraduate year 3), and expert (pediatric orthopedic attending) performance. With the "good" technique, mean peak temperatures were 43°C + 4.3°C. The highest recorded was 51.9°C. With the "poor" technique, mean peak temperature was 75.2°C + 17.3°C. The maximum temperature recorded with the "poor" technique was 112.4°C. Construct validity testing showed that novices had the highest increases in temperatures (12.9°C). There was a decline in heat generation as experience increased with the intermediate group (9.7°C), and the lowest heat generation was seen in the expert group (5.0°C). A novel task simulator and curriculum have been developed to assess competency and enhance performance in the application and removal of casts. There was a 32.2°C temperature decrease when the proper cast saw technique was used. Furthermore, the "poor" technique consistently achieved temperatures that would cause epidermal necrosis in patients. Clinical experience was a

  20. Modular Micromachined Si Heat Removal (MOMS Heat Removal): Electronic Integration and System Test

    National Research Council Canada - National Science Library

    Brown, Elliott

    2003-01-01

    ...: (1) insulated-gated bipolar transistors (IGBTs), and (2) laterally-diffused (LD) MOSFETs. Heat pipes were found to provide little or no advantage over conventional copper-based heat spreaders in both device applications...

  1. Treatment techniques for the removal of radioactive contaminants from drinking water

    International Nuclear Information System (INIS)

    Logsdon, Gary S.

    1978-01-01

    Maximum contaminant levels have been set for radioactive contaminants, as required by the Safe Drinking Water Act (PL 93-523). Treatment techniques are available for removing radium and beta-gamma emitters. Presently-used methods of removing radium-226 are precipitative lime softening (80-90% removal) ion exchange softening (95% removal) and reverse osmosis (95% removal). The 5 p Ci/l limit for radium can be met with conventional technology for raw waters in the 5-100 p Ci/l concentration range. Treatment for removal of beta or gamma emitters must be based upon chemical rather than radioactive characteristics of the contaminant. Reverse osmosis can remove a broad spectrum of ions and molecules from water, so it is the process most likely to be used. The maximum contaminant level for beta and gamma radioactivity is an annual dose equivalent to the total body or any organ not to exceed 4 m rem/year. The fate of radionuclides after removal from drinking water should be considered. Presently radium is disposed with other process wastes at softening plants removing radium. Confinement and disposal as a radioactive waste would be very expensive. (author)

  2. Failure Modes and Effects Analysis (FMEA) of the Residual Heat Removal System

    International Nuclear Information System (INIS)

    Eggleston, F.T.

    1976-01-01

    The Residual Heat Removal System (RHRS) transfer heat from the Reactor Coolant System (RCS) to the reactor plant Component Cooling System (CCS) to reduce the temperature of the RCS at a controlled rate during the second part of normal plant cooldown and maintains the desired temperature until the plant is restarted. By the use of an analytic tool, the Failure Modes and Effects Analysis, it is shown that the RHRS, because of its redundant two train design, is able to accommodate any credible component single failure with the only effect being an extension in the required cooldown time, thus demonstrating the reliability of the RHRS to perform its intended function

  3. System Study: Residual Heat Removal 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-12-01

    This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and post-trip shutdown-cooling) at 104 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trends were identified in the RHR results. A highly statistically significant decreasing trend was observed for the RHR injection mode start-only unreliability. Statistically significant decreasing trends were observed for RHR shutdown cooling mode start-only unreliability and RHR shutdown cooling model 24-hour unreliability.

  4. Method and device to remove the decay heat produced in the core of a nuclear reactor

    International Nuclear Information System (INIS)

    Loimann, E.; Reutler, H.

    1977-01-01

    For decay haet removal of the HTGR the heat absorbed by the top reflector is discharged by means of heat exchangers. For this purpose the heat exchangers are arranged between the top bricks consisting of graphite blocks. By convection or forced circulation with the aid of pumps the liquid coolant is flowing in a cycle between the individual heat exchangers connected in parallel and a heat sink arranged outside the containment. The distributing and collection pipes are mounted between the upper and lower thermal shield. The heat exchanger compartments themselves consist of double-walled hollow bodies with a disc-shaped section and a columnar part extending from there to one side respectively. (RW) [de

  5. Design of passive decay heat removal system using thermosyphon for low temperature and low pressure pool type LWR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; You, Byung Hyun; Jung, Yong Hun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    In seawater desalination process which doesn't need high temperature steam, the reactor has profitability. KAIST has be developing the new reactor design, AHR400, for only desalination. For maximizing safety, the reactor requires passive decay heat removal system. In many nuclear reactors, DHR system is loop form. The DHR system can be designed simple by applying conventional thermosyphon, which is fully passive device, shows high heat transfer performance and simple structure. DHR system utilizes conventional thermosyphon and its heat transfer characteristics are analyzed for AHR400. For maximizing safety of the reactor, passive decay heat removal system are prepared. Thermosyphon is useful device for DHR system of low pressure and low temperature pool type reactor. Thermosyphon is operated fully passive and has simple structure. Bundle of thermosyphon get the goal to prohibit boiling in reactor and high pressure in reactor vessel.

  6. Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

    International Nuclear Information System (INIS)

    Venker, Jeanne

    2015-01-01

    The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO 2 ) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO 2 Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO 2 Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO 2 Brayton cycles and to evaluate the introduced heat removal system

  7. Removing chromium and lead metals using phytoremediation technique

    Directory of Open Access Journals (Sweden)

    Al-Anbari Riyad

    2018-01-01

    Full Text Available Phytoremediation technique uses plants parts to remove, extract, and absorb heavy or toxic matter from soil and water. In the present study, Catharanthusroseus (Periwinkle and Nerium Oleander (Oleander were used for removing Chromium (Cr and Lead (Pb metals. These plant species were seeded in polyethylene pots containing 8kg of soil. Each pot was irrigated with wastewater for four months (May, June, July and August and accumulation of the considered metals was analyzed after every month for leaf, stem and root by using Atomic Absorption Spectrophotometer (AAS. This experimental work was carried out in the laboratories of Water Desalination Researches Unit - Building and Construction Engineering Department and Environmental Research Centre at the University of Technology in Baghdad City, Iraq. The concentration of Cr was found to be increased with time. High Cr concentration, 20.34 mg/kg, was recorded at August in leaf of Periwinkle and 19.61 mg/kg in root of Oleander in case of using 100% wastewater (WW. While, for Pb, the maximum concentration, 22 mg/kg, was recorded in June in leaf of Periwinkle and 19.5 mg/kg in steam of Oleander. Accordingly, Oleander has the maximum removal efficiency.

  8. Removal of the codeposited carbon layer using He-O glow discharge

    International Nuclear Information System (INIS)

    Kunz, C.L.; Causey, R.A.; Clift, M.; Wampler, W.R.; Cowgill, D.F.

    2007-01-01

    In this study we examine the combination of a He-O glow discharge with heating as a possible technique to remove deuterium from TFTR tiles. Samples were cut from a relatively large area containing a uniform codeposited layer of deuterium and carbon. Auger/SEM was used to generate micrographs of each of the samples. The samples were also examined using Rutherford backscattering to determine the near surface composition. Individual samples were then exposed to a He-O glow discharge while being heated. After the exposure, the samples were returned for Auger/SEM and RBS of the same areas examined prior to the exposure. Comparing the samples before and after exposure revealed that the amount of the codeposited layer removed was significantly less than 1 μm. Removal rates this low would suggest that He-O glow discharge with heating is insufficient to remove the thick layers predicted for ITER in a timely fashion

  9. Experimental and analytical studies on the passive residual heat removal system for the advanced integral type reactor

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki

    2004-01-01

    An experiment on the thermal-hydraulic characteristics of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART-P, has been performed, and its experimental results have been analyzed using a best-estimated system analysis code, MARS. The experiment is performed to investigate the performance of the passive residual heat removal system using the high temperature and high pressure thermal-hydraulic test facility (VISTA) which simulates the SMART-P. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfer through the PRHRS heat exchanger in the emergency cooldown tank is sufficient enough to enable a natural circulation of the coolant. Analysis on a typical PRHRS test has been carried out using the MARS code. The overall trends of the calculated flow rate, pressure, temperature, and heat transfer rate in the PRHRS are similar to the experimental data. There is good agreement between the experimental data and the calculated one for the fluid temperature in the PRHRS steam line. However, the calculated fluid temperature in the PRHRS condensate line is higher, the calculated coolant outlet temperature is lower, and the heat transfer rate through the PRHRS heat exchanger is lower than the experimental data. It seems that it is due to an insufficient heat transfer modeling in the pool such as the emergency cooldown tank in the MARS calculation. (author)

  10. Evaluation on the heat removal capacity of the first wall for water cooled breeder blanket of CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng, E-mail: jiangkecheng@ipp.ac.cn; Cheng, Xiaoman; Chen, Lei; Huang, Kai; Ma, Xuebin; Liu, Songlin

    2016-02-15

    Highlights: • Heat removal capacity of the FW is evaluated under BWR, PWR and He coolant inlet conditions. • Heat transfer property of the gas–liquid two phase and the two boiling crises are analyzed. • Heat removal capacity of water is larger than helium coolant. - Abstract: The water cooled ceramic breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). As an important component of the blanket, the FW should satisfy with the thermal requirements in any case. In this paper, three parameters including the heat removal capacity, coolant pressure drop as well as the temperature rise of the FW were investigated under different coolant velocity and heat flux from the plasma. Using the same first wall structure, two main water cooled schemes including Boiling Water Reactor (BWR, 7 MPa pressure and 265 °C temperature inlet) and Pressurized Water Reactor (PWR, 15 MPa pressure and 285 °C temperature inlet) conditions are discussed in the thermal hydraulic calculation. For further research, the thermal hydraulic characteristics of using helium as coolant (8 MPa pressure, 300 °C temperature inlet) are also explored to provide CFETR blanket design with more useful data supports. Without regard to the outlet coolant condition requirements of the blanket, the results indicate that the ultimate heat flux that the FW can resist is 2.2 MW/m{sup 2} at velocity of 5 m/s for BWR, 2.0 MW/m{sup 2} at velocity of 5 m/s for PWR and 0.87 MW/m{sup 2} for helium at velocity 100 m/s under the chosen operation condition. The detrimental departure from nucleate boiling (DNB) crisis would occur at the velocity of 1 m/s under the heat flux of 3 MW/m{sup 2} and dry out crisis appears at the velocity of less than 0.2 m/s with the heat flux of more than 1 MW/m{sup 2} for BWR. The further blanket/FW optimization design is provided with more useful data references according to the abundant calculation results.

  11. Control of reactor coolant flow path during reactor decay heat removal

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  12. Development of evaluation method for heat removal design of dry storage facilities. Pt. 4. Numerical analysis on vault storage system of cross flow type

    International Nuclear Information System (INIS)

    Sakamoto, Kazuaki; Hattori, Yasuo; Koga, Tomonari; Wataru, Masumi

    1999-01-01

    On the basis of the result of the heat removal test on vault storage system of cross flow type using the 1/5 scale model, an evaluation method for the heat removal design was established. It was composed of the numerical analysis for the convection phenomena of air flow inside the whole facility and that for the natural convection and the detailed turbulent mechanism near the surface of the storage tube. In the former analysis, air temperature distribution in the storage area obtained by the calculation gave good agreement within ±3degC with the test result. And fine turbulence models were introduced in the latter analysis to predict the separation flow in the boundary layer near the surface of the storage tube and the buoyant flow generated by the heat from the storage tube. Furthermore, the properties of removing the heat in a designed full-scale storage facility, such as flow pattern in the storage area, temperature and heat transfer rate of the storage tubes, were evaluated by using each of three methods, which were the established numerical analysis method, the experimental formula demonstrated in the heat removal test and the conventional evaluation method applied to the past heat removal design. As a result, the safety margin and issues included in the methods were grasped, and the measures to make a design more rational were proposed. (author)

  13. Simultaneous removal of NO and SO2 using vacuum ultraviolet light (VUV)/heat/peroxymonosulfate (PMS).

    Science.gov (United States)

    Liu, Yangxian; Wang, Yan; Wang, Qian; Pan, Jianfeng; Zhang, Jun

    2018-01-01

    Simultaneous removal process of SO 2 and NO from flue gas using vacuum ultraviolet light (VUV)/heat/peroxymonosulfate (PMS) in a VUV spraying reactor was proposed. The key influencing factors, active species, reaction products and mechanism of SO 2 and NO simultaneous removal were investigated. The results show that vacuum ultraviolet light (185 nm) achieves the highest NO removal efficiency and yield of and under the same test conditions. NO removal is enhanced at higher PMS concentration, light intensity and oxygen concentration, and is inhibited at higher NO concentration, SO 2 concentration and solution pH. Solution temperature has a double impact on NO removal. CO 2 concentration has no obvious effect on NO removal. and produced from VUV-activation of PMS play a leading role in NO removal. O 3 and ·O produced from VUV-activation of O 2 also play an important role in NO removal. SO 2 achieves complete removal under all experimental conditions due to its very high solubility in water and good reactivity. The highest simultaneous removal efficiency of SO 2 and NO reaches 100% and 91.3%, respectively. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. Removal of benzaldehyde from a water/ethanol mixture by applying scavenging techniques

    DEFF Research Database (Denmark)

    Mitic, Aleksandar; Skov, Thomas; Gernaey, Krist V.

    2017-01-01

    A presence of carbonyl compounds is very common in the food industry. The nature of such compounds is to be reactive and thus many products involve aldehydes/ketones in their synthetic routes. By contrast, the high reactivity of carbonyl compounds could also lead to formation of undesired compounds......, such as genotoxic impurities. It can therefore be important to remove carbonyl compounds by implementing suitable removal techniques, with the aim of protecting final product quality. This work is focused on benzaldehyde as a model component, studying its removal from a water/ethanol mixture by applying different...

  15. Assessment of Haar Wavelet-Quasilinearization Technique in Heat Convection-Radiation Equations

    Directory of Open Access Journals (Sweden)

    Umer Saeed

    2014-01-01

    Full Text Available We showed that solutions by the Haar wavelet-quasilinearization technique for the two problems, namely, (i temperature distribution equation in lumped system of combined convection-radiation in a slab made of materials with variable thermal conductivity and (ii cooling of a lumped system by combined convection and radiation are strongly reliable and also more accurate than the other numerical methods and are in good agreement with exact solution. According to the Haar wavelet-quasilinearization technique, we convert the nonlinear heat transfer equation to linear discretized equation with the help of quasilinearization technique and apply the Haar wavelet method at each iteration of quasilinearization technique to get the solution. The main aim of present work is to show the reliability of the Haar wavelet-quasilinearization technique for heat transfer equations.

  16. Numerical analyses of the effect of a biphasic thermosyphon vapor channel sizes on the heat transfer intensity when heat removing from a power transformer of combined heat and power station

    Directory of Open Access Journals (Sweden)

    Nurpeiis Atlant

    2017-01-01

    Full Text Available Numerical analyses of the effect of a biphasic thermosyphon vapor channel sizes on the heat transfer intensity was conducted when heat removing from an oil tank of a power transformer of combined heat and power station (CHP. The power transformer cooling system by the closed biphasic thermosyphon was proposed. The mathematical modeling of heat transfer and phase transitions of coolant in the thermosyphon was performed. The problem of heat transfer is formulated in dimensionless variables “velocity vorticity vector – current function – temperature” and solved by finite difference method. As a result of numerical simulation it is found that an increase in the vapor channel length from 0.15m to 1m leads to increasing the temperature difference by 3.5 K.

  17. Numerical simulation of flow field in cooling tower of passive residual heat removal system of HTGR

    International Nuclear Information System (INIS)

    Li Xiaowei; Zhang Li; Wu Xinxin; He Shuyan

    2011-01-01

    Environmental wind will influence the working conditions of natural convection cooling tower. The velocity and temperature fields in the natural convection cooling tower of the HTGR residual heat removal system at different environmental wind velocities were numerically simulated. The results show that, if there is no wind baffle, the flow in the cooling tower is blocked when environmental wind velocity is higher than 6 m/s, residual heat can hardly be removed, and when wind velocity is higher than 9 m/s, the air even flow downwards in the tower, so wind baffle is very necessary. With the wind baffle installed, the cooling tower works well at the wind speed even higher than 9 m/s. The optimum baffle size and positions are also analyzed. (authors)

  18. Reliability analysis on passive residual heat removal of AP1000 based on Grey model

    Energy Technology Data Exchange (ETDEWEB)

    Qi, Shi; Zhou, Tao; Shahzad, Muhammad Ali; Li, Yu [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, Beijing (China); Jiang, Guangming [Nuclear Power Institute of China, Chengdu (China). Science and Technology on Reactor System Design Technology Laboratory

    2017-06-15

    It is common to base the design of passive systems on the natural laws of physics, such as gravity, heat conduction, inertia. For AP1000, a generation-III reactor, such systems have an inherent safety associated with them due to the simplicity of their structures. However, there is a fairly large amount of uncertainty in the operating conditions of these passive safety systems. In some cases, a small deviation in the design or operating conditions can affect the function of the system. The reliability of the passive residual heat removal is analysed.

  19. Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Venker, Jeanne

    2015-03-31

    The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO{sub 2}) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO{sub 2} Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO{sub 2} Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO{sub 2} Brayton cycles and to evaluate the introduced

  20. A value/impact assessment for alternative decay heat removal systems

    International Nuclear Information System (INIS)

    Cave, L.; Kastenberg, W.E.; Lin, K.Y.

    1984-01-01

    A Value/Impact assessment for several alternative decay heat removal systems has been carried out using several measures. The assessment is based on an extension of the methodology presented in the Value/Impact Handbook and includes the effects of uncertainty. The assessment was carried out as a function of site population density, existing plant features, and new plant features. Value/Impact measures based on population dose are shown to be sensitive to site, while measures which monetize and aggregate risk are less so. The latter are dominated by on-site costs such as replacement power costs. (orig.)

  1. Radon-removal techniques for small community public water supplies

    International Nuclear Information System (INIS)

    Kinner, N.E.; Malley, J.P.; Clement, J.A.; Quern, P.A.; Schell, G.S.

    1990-08-01

    The report presents the results of an evaluation, performed by the University of New Hampshire--Environmental Research Group (ERG), of radon removal in small community water supplies using full-scale granular activated carbon adsorption, diffused bubble aeration and packed tower aeration. Various low technology alternatives, such as loss in a distribution system and addition of coarse bubble aeration to a pilot-scale atmospheric storage tank were also evaluated. The report discusses each of the treatment alternatives with respect to their radon removal efficiency, potential problems (i.e., waste disposal, radiation exposure and intermedia pollution), and economics in small community applications. In addition, several sampling methods, storage times, scintillation cocktails and extraction procedures currently used in the liquid scintillation technique for analysis of radon in water were compared

  2. Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

    Energy Technology Data Exchange (ETDEWEB)

    De Wild, P.J.; Nyqvist, R.G.; De Bruijn, F.A.; Stobbe, E.R. [ECN Hydrogen and Clean Fossil Fuels, Petten (Netherlands)

    2006-02-15

    Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC)) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20 ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG.

  3. Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

    Energy Technology Data Exchange (ETDEWEB)

    de Wild, P.J.; Nyqvist, R.G.; de Bruijn, F.A.; Stobbe, E.R. [Energy Research Centre of The Netherlands ECN, P.O. Box 1, 1755 ZG Petten (Netherlands)

    2006-09-22

    Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG. (author)

  4. Development of margin assessment methodology of decay heat removal function against external hazards. (2) Tornado PRA methodology

    International Nuclear Information System (INIS)

    Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

    2014-01-01

    Probabilistic Risk Assessment (PRA) for external events has been recognized as an important safety assessment method after the TEPCO's Fukushima Daiichi nuclear power station accident. The PRA should be performed not only for earthquake and tsunami which are especially key events in Japan, but also the PRA methodology should be developed for the other external hazards (e.g. tornado). In this study, the methodology was developed for Sodium-cooled Fast Reactors paying attention to that the ambient air is their final heat sink for removing decay heat under accident conditions. First, tornado hazard curve was estimated by using data recorded in Japan. Second, important structures and components for decay heat removal were identified and an event tree resulting in core damage was developed in terms of wind load and missiles (i.e. steel pipes, boards and cars) caused by a tornado. Main damage cause for important structures and components is the missiles and the tornado missiles that can reach those components and structures placed on high elevations were identified, and the failure probabilities of the components and structures against the tornado missiles were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or outtake in the decay heat removal system, and a probability of failure caused by the missile impacts. Finally, the event tree was quantified. As a result, the core damage frequency was enough lower than 10 -10 /ry. (author)

  5. Study on grey theoretical model of passive residual heat removal system

    International Nuclear Information System (INIS)

    Zhou Tao; Yang Ruichang; Su, G.H.; Jia Dounan; Sugiyama, K.

    2004-01-01

    Natural Circulation Passive Residual Heat Removal System is treated as a Grey System by taking into account of its complexity and uncertainty of effect for factors each other. The magnitude and degree of some factors are confirmed by grey incidence analysis method; The one-one relationship of some variables is built by GM (1, 1) model; The relationship between key factor and other effect factors is built (1, 4) model. Grey model shows its more advantage of precision through comparing with multivariate model. (author)

  6. Influence of wick properties in a vertical LHP on remove waste heat from electronic equipment

    International Nuclear Information System (INIS)

    Smitka, Martin; Nemec, Patrik; Malcho, Milan

    2014-01-01

    The loop heat pipe is a vapour-liquid phase-change device that transfers heat from evaporator to condenser. One of the most important parts of the LHP is the porous wick structure. The wick structure provides capillary force to circulate the working fluid. To achieve good thermal performance of LHP, capillary wicks with high permeability and porosity and fine pore radius are expected. The aim of this work is to develop porous wick of sintered nickel powder with different grain sizes. These porous wicks were used in LHP and there were performed a series of measurements to remove waste heat from the insulated gate bipolar transistor (IGBT)

  7. Influence of wick properties in a vertical LHP on remove waste heat from electronic equipment

    Energy Technology Data Exchange (ETDEWEB)

    Smitka, Martin, E-mail: martin.smitka@fstroj.uniza.sk, E-mail: patrik.nemec@fstroj.uniza.sk, E-mail: milan.malcho@fstroj.uniza.sk; Nemec, Patrik, E-mail: martin.smitka@fstroj.uniza.sk, E-mail: patrik.nemec@fstroj.uniza.sk, E-mail: milan.malcho@fstroj.uniza.sk; Malcho, Milan, E-mail: martin.smitka@fstroj.uniza.sk, E-mail: patrik.nemec@fstroj.uniza.sk, E-mail: milan.malcho@fstroj.uniza.sk [University of Žilina, Faculty of Mechanical Engineering, Department of Power Engeneering, Univerzitna 1, 010 26 Žilina (Slovakia)

    2014-08-06

    The loop heat pipe is a vapour-liquid phase-change device that transfers heat from evaporator to condenser. One of the most important parts of the LHP is the porous wick structure. The wick structure provides capillary force to circulate the working fluid. To achieve good thermal performance of LHP, capillary wicks with high permeability and porosity and fine pore radius are expected. The aim of this work is to develop porous wick of sintered nickel powder with different grain sizes. These porous wicks were used in LHP and there were performed a series of measurements to remove waste heat from the insulated gate bipolar transistor (IGBT)

  8. Effect of short-term heat acclimation training on kinetics of lactate removal following maximal exercise.

    Science.gov (United States)

    Dileo, Tsavis D; Powell, Jeffrey B; Kang, Hyoung K; Roberge, Raymond J; Coca, Aitor; Kim, Jung-Hyun

    2016-01-01

    Heat acclimation (HA) evokes numerous physiological adaptations, improves heat tolerance and has also been shown to enhance lactate (LA) responses during exercise, similar to that seen with endurance training. The purpose of this study was to examine whether HA improves the body's ability to remove LA during recovery following maximal exercise. Ten healthy men completed two trials of maximal treadmill exercise (pre- and post-HA) separated by 5 days of HA. Each day of HA consisted of two 45 minute periods of cycling at ~50% VO2max separated by a 15min rest period in an environmental chamber (T(db) 45° C, RH 20%). In pre-/post-HA trials, venous blood was collected during 60 minutes of recovery to determine LA concentrations and removal kinetics (A2: amplitude and y2: velocity constant) using bi-exponential curve fitting. Physiological adaptation to heat was significantly developed during HA, as evidenced by end-exercise T(re) (DAY1 vs. 5) (38.89±0.56 vs. 38.66±0.44° C), T(sk) (38.07±0.51 vs. 37.66±0.48° C), HR (175.0±9.9 vs. 165.0±18.5 beats·min(-1)), and sweat rate (1.24 ±.26 vs. 1.47 ±0.27 L·min(-1)) (PLA concentrations (LA(0min): 8.78±1.08 vs. 8.69±1.23; LA(peak): 10.97±1.77 vs. 10.95±1.46; and La(60min); 2.88±0.82 vs. 2.96±0.93 mmol·L(-1)) or removal kinetics (A2: -13.05±7.05 vs -15.59±7.90 mmol.L(-1) and y2: 0.02±0.01 vs. 0.03±.01 min(-1)). The present study concluded that, while effective in inducing thermo-physiological adaptations to heat stress, short-term HA does not improve the body's ability to remove LA following maximal exercise. Therefore, athletes and workers seeking faster LA recovery from intense physical activity may not benefit from short-term HA.

  9. Laparoscopic Cornuectomy as a Technique for Removal of Essure Microinserts.

    Science.gov (United States)

    Thiel, Luke; Rattray, Darrien; Thiel, John

    2017-01-01

    The authors present a laparoscopic technique for complete removal of Essure microinserts (including nitinol coil and positron emission tomography fibers). Step-wise instruction using video. The study was granted a Research Ethics Board exemption because the Regina Qu'Appelle Health Region Research Ethics Board does not require ethics board approval for single case submissions. Tertiary care hospital. Patient requesting removal of Essure inserts because of post-placement discomfort. Recent concern regarding adverse outcomes (persistent pelvic pain, device malposition, nickel allergy) after Essure placement has led to a small percentage of women requesting removal of the coils. Laparoscopic salpingectomy and salpingostomy have been successfully used for removal. Hysteroscopic removal has been achieved up to 6 weeks after placement; however, because of the fibrosis-inducing mechanism of the inserts, there is theoretical concern regarding fragmentation or incomplete removal with a cut and pull approach. The authors used a laparoscopic surgical approach for removal of the Essure microinserts "en bloc" by performing a salpingectomy and mini-resection of the uterine cornua to the level of the endometrium. This approach ensures complete extraction of the Essure microinserts. The surgery was completed in a tertiary care hospital operating theatre with standard laparoscopic and electrosurgical instruments using a 10-mm infraumbilical port and two 5-mm ports in the left lower quadrant. En bloc resection of the fallopian tubes, uterine cornua, and Essure microinserts is a feasible laparoscopic approach to ensure complete removal of Essure microinserts. This approach is technically straightforward and can be achieved with minimal blood loss. Copyright © 2016 AAGL. Published by Elsevier Inc. All rights reserved.

  10. Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

    Energy Technology Data Exchange (ETDEWEB)

    Li, Cheng, E-mail: 510395453@qq.com [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China); Li, Le [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Junming [Tsinghua University, Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Thermal Engineering, Beijing 100084 (China); Zhang, Yajun [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Zhihui [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China)

    2017-03-15

    Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

  11. Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

    International Nuclear Information System (INIS)

    Li, Cheng; Li, Le; Li, Junming; Zhang, Yajun; Li, Zhihui

    2017-01-01

    Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

  12. Model techniques for testing heated concrete structures

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1983-01-01

    Experimental techniques are described which may be used in the laboratory to measure strains of model concrete structures representing to scale actual structures of any shape or geometry, operating at elevated temperatures, for which time-dependent creep and shrinkage strains are dominant. These strains could be used to assess the distribution of stress in the scaled structure and hence to predict the actual behaviour of concrete structures used in nuclear power stations. Similar techniques have been employed in an investigation to measure elastic, thermal, creep and shrinkage strains in heated concrete models representing to scale parts of prestressed concrete pressure vessels for nuclear reactors. (author)

  13. Control of reactor coolant flow path during reactor decay heat removal

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  14. "Push back" technique: A simple method to remove broken drill bit from the proximal femur.

    Science.gov (United States)

    Chouhan, Devendra K; Sharma, Siddhartha

    2015-11-18

    Broken drill bits can be difficult to remove from the proximal femur and may necessitate additional surgical exploration or special instrumentation. We present a simple technique to remove a broken drill bit that does not require any special instrumentation and can be accomplished through the existing incision. This technique is useful for those cases where the length of the broken drill bit is greater than the diameter of the bone.

  15. Safe Removal of an Encrusted Nephrostomy Tube Using a Vascular Sheath: A Technique Revisited

    Energy Technology Data Exchange (ETDEWEB)

    Farooq, Ammad, E-mail: faroamm@aol.com; Agarwal, Sanjay; Jones, Vaughan [Wrexham Maelor Hospital, Department of Radiology (United Kingdom)

    2013-06-15

    With the advent of interventional radiology and the decrease in mortality from chronic ailments, especially malignancy, percutaneous nephrostomy has become a commonly used safe technique for temporary relief of renal tract obstruction or for urinary diversion. However, these are associated with risks of infection, particularly septicaemia, colonisation, and blockage. Another significant complication is difficulty in removal due to encrustation. We describe a useful technique used in our department for the past few years and cite four cases of variable presentation and complexity for removal of an encrusted nephrostomy tube. No mention of this technique was found recent literature. An almost similar technique was described in the 1980s ''Pollack and Banner (Radiology 145:203-205, 1982), Baron and McClennan (Radiology 141:824, 1981)''. It is possible that experienced operators may have used this technique. We revisit it with pictographic representation, describing its use with currently available equipment, for benefit of operators who are not aware of this technique.

  16. SASSYS validation with the EBR-II shutdown heat removal tests

    International Nuclear Information System (INIS)

    Herzog, J.P.

    1989-01-01

    SASSYS is a coupled neutronic and thermal hydraulic code developed for the analysis of transients in liquid metal cooled reactors (LMRs). The code is especially suited for evaluating of normal reactor transients -- protected (design basis) and unprotected (anticipated transient without scram) transients. Because SASSYS is heavily used in support of the IFR concept and of innovative LMR designs, such as PRISM, a strong validation base for the code must exist. Part of the validation process for SASSYS is analysis of experiments performed on operating reactors, such as the metal fueled Experimental Breeder Reactor -- II (EBR-II). During the course of a series of historic whole-plant experiments, EBR-II illustrated key safety features of metal fueled LMRs. These experiments, the Shutdown Heat Removal Tests (SHRT), culminated in unprotected loss of flow and loss of heat sink transients from full power and flow. Analysis of these and earlier SHRT experiments constitutes a vital part of SASSYS validation, because it facilitates scrutiny of specific SASSYS models and of integrated code capability. 12 refs., 11 figs

  17. Experimental data processing technique for nonstationary heat transfer on fuel rod simulators

    International Nuclear Information System (INIS)

    Nikonov, S.P.; Nikonov, A.P.; Belyukin, V.A.

    1982-01-01

    Non-stationary heat-transfer data processing is considered in connection with experimental studies of the emergency cooling whereat fuel rod imitators both with direct and indirect shell heating were used. The objective of data processing was obtaining the temperature distribution within the imitator, the heat flux removed by the coolant and the shell-coolant heat-transfer coefficient. The special attention was paid to the temperature distribution calculation at the data processing during the reflooding experiments. In this case two factors are assumed to be known: the time dependency of temperature variation at a certain point within the imitator cross-section and the heat flux at some point of the same cross-section. The initial data preparation for calculations, employing the procedure of smoothing by cubic spline functions, is considered as well, with application of an algorithm reported in the literature, which is efficient for the given functional dependency wherein the deviation in each point is known [ru

  18. Design and transient analyses of emergency passive residual heat removal system of CPR1000

    International Nuclear Information System (INIS)

    Zhang, Y.P.; Qiu, S.Z.; Su, G.H.; Tian, W.X.

    2012-01-01

    Highlights: ► Designing an EPRHRs for CPR1000. ► Developing a RELAP model of the EPRHRs. ► The EPRHRs could take away the decay heat effectively. - Abstract: The steam generator secondary emergency passive residual heat removal system (EPRHRs) is a new design for traditional generation II + reactor CPR1000. The EPRHRs is designed to improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the station blackout or loss of heat sink accident. The EPRHRs consists of steam generator (SG), heat exchanger (HX), emergency makeup tank (EMT), cooling water tank (CWT), and corresponding pipes and valves. In order to improve the safety and reliability of CPR1000, the model of the primary loop and the EPRHRs was developed to investigate residual heat removal capability of the EPRHRs and the transient characteristics of the primary loop affected by the EPRHRs using RELAP5/MOD3.4. The transient characteristics of the primary loop and the EPRHRs were calculated in the event of station blackout accident. Sensitivity studies of the EPRHRs were also conducted to investigate the response of the primary loop and the EPRHRs on the main parameters of the EPRHRs. The EPRHRs could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRs could take away the decay heat from the primary loop effectively, and that the single-phase and two-phase natural circulations were established in the primary loop and EPRHRs loop, respectively. The results also indicated that the effect of isolation valve open time on the transient characteristics of the primary loop was little. However, the effect of isolation valve open time on the EPRHRs condensate flow was relatively greater. The isolation valves should not be opened too rapidly during the isolation valve opening process, and the isolation valve opening time should be greater than 10 s, which could avoid the

  19. Heat exchanger device and method for heat removal or transfer

    Science.gov (United States)

    Koplow, Jeffrey P

    2013-12-10

    Systems and methods for a forced-convection heat exchanger are provided. In one embodiment, heat is transferred to or from a thermal load in thermal contact with a heat conducting structure, across a narrow air gap, to a rotating heat transfer structure immersed in a surrounding medium such as air.

  20. GOTHIC-IST 6.1b code validation exercises relating to heat removal by dousing and air coolers in CANDU containment

    International Nuclear Information System (INIS)

    Ramachandran, S.; Krause, M.; Nguyen, T.

    2003-01-01

    This paper presents validation results relating to the use of the GOTHIC containment analysis code for CANDU safety analysis. The validation results indicate that GOTHIC predicts heat removal by dousing and air cooler heat transfer with reasonable accuracy. (author)

  1. Removal of a Broken Cannulated Intramedullary Nail: Review of the Literature and a Case Report of a New Technique

    Directory of Open Access Journals (Sweden)

    Amr A. Abdelgawad

    2013-01-01

    Full Text Available Nonunion of long bones fixed with nails may result in implant failure. Removal of a broken intramedullary nail may be a real challenge. Many methods have been described to allow for removal of the broken piece of the nail. In this paper, we are reviewing the different techniques to extract a broken nail, classifying them into different subsets, and describing a new technique that we used to remove a broken tibial nail with narrow canal. Eight different categories of implant removal methods were described, with different methods within each category. This classification is very comprehensive and was never described before. We described a new technique (hook captured in the medulla by flexible nail introduced from the locking hole which is a valuable technique in cases of nail of a small diameter where other methods cannot be used because of the narrow canal of the nail. Our eight categories for broken nail removal methods simplify the concepts of nail removal and allow the surgeon to better plan for the removal procedure.

  2. Removing a tick: proper technique in children

    Directory of Open Access Journals (Sweden)

    Susana Rueda Pérez

    2012-06-01

    Full Text Available There is medical consensus on the need to remove the tick within 24 hours the mite parasites to the human host, to avoid possible complications. The preferred way is by gently traction the mite, aided by forceps without twisting or chokes with toxic agents, because of the possibility that the mite excretes bacteria mixed with substances. The average time of extraction is estimated between one or three minutes. In children parasitized by ticks this amount of time can be excessive when it’s necessary restraint without the consent of the minor. Using this technique we reduce the time to seconds and the damage caused to the skin is minimal.

  3. Removal of Fe(II) from tap water by electrocoagulation technique

    International Nuclear Information System (INIS)

    Ghosh, D.; Solanki, H.; Purkait, M.K.

    2008-01-01

    Electrocoagulation (EC) is a promising electrochemical technique for water treatment. In this work electrocoagulation (with aluminum as electrodes) was studied for iron Fe(II) removal from aqueous medium. Different concentration of Fe(II) solution in tap water was considered for the experiment. During EC process, various amorphous aluminum hydroxides complexes with high sorption capacity were formed. The removal of Fe(II) was consisted of two principal steps; (a) oxidation of Fe(II) to Fe(III) and (b) subsequent removal of Fe(III) by the freshly formed aluminum hydroxides complexes by adsorption/surface complexation followed by precipitation. Experiments were carried out with different current densities ranging from 0.01 to 0.04 A/m 2 . It was observed that the removal of Fe(II) increases with current densities. Inter electrode distance was varied from 0.005 to 0.02 m and was found that least inter electrode distance is suitable in order to achieve higher Fe(II) removal. Other parameters such as conductivity, pH and salt concentration were kept constant as per tap water quality. Satisfactory iron removal of around 99.2% was obtained at the end of 35 min of operation from the initial concentration of 25 mg/L Fe(II). Iron concentration in the solution was determined using Atomic absorption spectrophotometer. By products obtained from the electrocoagulation bath were analyzed by SEM image and corresponding elemental analysis (EDAX). Cost estimation for the electrocoagulation was adopted and explained well. Up to 15 mg/L of initial Fe(II) concentration, the optimum total cost was 6.05 US$/m 3 . The EC process for removing Fe(II) from tap water is expected to be adaptable for household use

  4. Investigation on natural convection decay heat removal for the EFR status of the program

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, F [Kernforschungszentrum Karlsruhe (Germany); Essig, C [Siemens AG, Bergisch Gladbach (Germany); Georgeoura, S [AEA Reactor Service, Dounreay (United Kingdom); Tenchine, D [CEA Grenoble (France)

    1993-02-01

    The European Research and Development (R+D) Program on decay heat removal by natural convection for the European Fast Reactor (EFR) covers the calculational methods and the model experiments performed for code validation. The studies concentrate on important physical effects of the cooling modes within the primary system and the direct reactor cooling circuits and include reactor experiments. (author)

  5. Investigation on natural convection decay heat removal for the EFR status of the program

    International Nuclear Information System (INIS)

    Hofmann, F.; Essig, C; Georgeoura, S.; Tenchine, D.

    1993-01-01

    The European Research and Development (R+D) Program on decay heat removal by natural convection for the European Fast Reactor (EFR) covers the calculational methods and the model experiments performed for code validation. The studies concentrate on important physical effects of the cooling modes within the primary system and the direct reactor cooling circuits and include reactor experiments. (author)

  6. Assessment of the advantages of a residual heat removal system inside the reactor pressure vessel

    International Nuclear Information System (INIS)

    Gautier, G.M.

    1995-01-01

    In the framework of research on diversified means for removing the residual heat from pressurized water reactors, the CEA is studying a passive system called RRP (Refroidissement du Reacteur au Primaire, or primary circuit cooling system), which includes integrated heat-exchangers and a layout of the internal structures so as to obtain convection from the primary circuit inside the vessel, whatever the state of the loops. This system is operational for all primary circuit temperatures and pressures, as well as for a wide range of conditions: it is independent of the state of the loops, even if the volume of water in the primary circuit is small, it is compatible with either a passive or an active operation mode, and compatible with any other decay heat removal systems. An evaluation is presented here of the performance of the RRP system in the event of a small primary circuit break in a totally passive operation mode without the intervention of another system. The results of this evaluation show the interest of such a system: a clear increase of the time-delay for the implementation of a low pressure safety injection system, no need for the use of a high pressure safety injection system. (author). 4 refs., 7 figs., 1 tab

  7. Assessment of the advantages of a residual heat removal system inside the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [Commissariat a l`Energie Atomique, Saint-Paul-Lez-Durance (France)

    1995-09-01

    In the framework of research on diversified means for removing residual heat from pressurized water reactors, the CEA is studying a passive system called RRP (Refroidissement du Reacteur au Primaire, or primary circuit cooling system). This system consists of integrated heat-exchangers and a layout of the internal structures so as to obtain convection from the primary circuit inside the vessel, whatever the state of the loops. This system is operational for all primary circuit temperatures and pressures, as well as for a wide range of conditions: such as independent from the state of the loops, low volume of water in the primary circuit, compatibility with either a passive or an active operation mode, and compatibility with any other decay heat removal systems. This paper presents an evaluation of the performance of the RRP system in the event of a small primary circuit break in a totally passive operation mode without the intervention of any another system. The results of this evaluation show the potential interest of such a system: a clear increase of the time-delay for the implementation of a low pressure safety injection system and no need for the use of a high pressure safety injection system.

  8. Non-aqueous removal of sodium from reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Welch, F H; Steele, O P [Rockwell International, Atomics International Division, Canoga Park (United States)

    1978-08-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component.

  9. Non-aqueous removal of sodium from reactor components

    International Nuclear Information System (INIS)

    Welch, F.H.; Steele, O.P.

    1978-01-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component

  10. Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

  11. Mitigation Measures Following a Loss-of-Residual-Heat-Removal Event During Shutdown

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Bang, Young Seok; Kim, Hho Jung

    2000-01-01

    The transient following a loss-of-residual-heat-removal event during shutdown was analyzed to determine the containment closure time (CCT) to prevent uncontrolled release of fission products and the gravity-injection path and rate (GIPR) for effective core cooling using the RELAP5/MOD3.2 code. The plant conditions of Yonggwang Units 3 and 4, a pressurized water reactor (PWR) of 2815-MW(thermal) power in Korea, were reviewed, and possible event sequences were identified. From the CCT analysis for the five cases of typical plant configurations, it was estimated for the earliest CCT to be 40 min after the event in a case with a large cold-leg opening and emptied steam generators (SGs). However, the case with water-filled SGs significantly delayed the CCT through the heat removal to the secondary side. From the GIPR analysis for the six possible gravity-injection paths from the refueling water storage tank (RWST), the case with the injection point and opening on the other leg side was estimated to be the most suitable path to avoid core boiling. In addition, from the sensitivity study, it was evaluated for the plant to be capable of providing the core cooling for the long-term transient if nominal RWST water is available. As a result, these analysis methods and results will provide useful information in understanding the plant behavior and preparing the mitigation measures after the event, especially for Combustion Engineering-type PWR plants. However, to directly apply the analysis results to the emergency procedure for such an event, additional case studies are needed for a wide range of operating conditions such as reactor coolant inventory, RWST water temperature, and core decay heat rate

  12. PANDA passive decay heat removal transient test results

    International Nuclear Information System (INIS)

    Bandurski, Th.; Dreier, J.; Huggenberger, M.

    1997-01-01

    PANDA is a large scale facility for investigating the long-term decay heat removal from the containment of a next generation of 'passive' Advanced Light Water Reactors (ALWR). PANDA was used to examine the long-term LOCA response of the Passive Containment Cooling System (PCCS) for the General Electric (GE) Simplified Boiling Water Reactor (SBWR). The first PANDA test series had the dual objectives of demonstrating the performance of the SBWR PCCS and extending the data base available for containment analysis code qualification. The test objectives also include the study of the effects of mixing and stratification of steam and noncondensible gases in the drywell (DW) and in the suppression chamber or wetwell (WW). Ten tests were conducted in the course of the PANDA SBWR Program. The tests demonstrated a favorable and robust overall PCCS performance under different conditions. The present paper focuses on the main phenomena observed during the tests with respect to PCCS operation and DW gas mixing. (author)

  13. Numerical calculation and analysis of natural convection removal of the spent fuel residual heat of 10 MW high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Wang Jinhua; Huang Yifan; Wu Bin

    2013-01-01

    The spent fuel of 10 MW High Temperature Gas Cooled Reactor (HTR-10) could be stored in the shielded tank, and the tank is stored in the concrete shielded canister in spent fuel storage room, the residual heat of the spent fuel could be removed by the air. The ability of residual heat removal is analyzed in the paper, and the temperature field is numerically calculated through FEA program ANSYS, the analysis and the calculation are used to validate the safety of the spent fuel and the tank, the ultimate temperature of the spent fuel and the tank should below the safety limit. The calculation shows that the maximum temperature locates in the middle of the fuel pebble bed in the spent fuel tank, and the temperature decreases gradually with radial distance, the temperature in the tank body is evenly distributed, and the temperature in the concrete shielded canister decreases gradually with radial distance. It is feasible to remove the residual heat of the spent fuel storage tank by natural ventilation, in natural ventilation condition, the temperature of the spent fuel and the tank is lower than the temperature limit, which provides theoretical evidence for the choice of the residual heat removal method. (authors)

  14. SASSYS-1 modelling of RVACS/RACS heat removal in an LMR

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1987-01-01

    The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This air-side RVACS/RACS model is coupled with the sodium-side primary loop thermal hydraulics model in SASSYS-1 to give a complete treatment of the problem. Application of this model to an unprotected loss-of-flow event in the PRISM rector shows that in the long run the RVACS cooling is sufficient to prevent unacceptably high system temperatures in this case

  15. Design and modeling of an advanced marine machinery system including waste heat recovery and removal of sulphur oxides

    DEFF Research Database (Denmark)

    Frimann Nielsen, Rasmus; Haglind, Fredrik; Larsen, Ulrik

    2013-01-01

    -stroke diesel engine and a conventional waste heat recovery system. The results suggest that an organic Rankine cycle placed after the conventional waste heat recovery system is able to extract the sulphuric acid from the exhaust gas, while at the same time increase power generation from waste heat by 32...... consists of a two-stroke diesel engine, the wet sulphuric process for sulphur removal and an advanced waste heat recovery system including a conventional steam Rankine cycle and an organic Rankine cycle. The results are compared with those of a state-of-the-art machinery system featuring a two...

  16. Study on thermalhydraulics of natural circulation decay heat removal in FBR. Experiment with water of typical reactor trip in the demonstration FBR

    International Nuclear Information System (INIS)

    Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

    2010-01-01

    Intending to enhance safety and to reduce costs, an FBR plant is being developed in Japan. In relies solely on natural circulation of the primary cooling loop to remove a decay heat of the core after reactor trips. A water test was carried out to advance the development. The test used a 1/10 reduced scale model simulating the core and cooling systems. The experiments simulated representative accidents from steady state to decay heat removal through reactor trip and clarified thermal-hydraulic issues on the thermal circulation performance. Some modifications of the system design were proposed for solving serious problems of natural circulation. An improved design complying with the suggestions will make it possible for natural circulation of the cooling systems to remove the decay heat of the core without causing and unstable or unpredictable change. (author)

  17. Shutdown decay heat removal analysis: Plant case studies and special issues: Summary report

    International Nuclear Information System (INIS)

    Ericson, D.M. Jr.; Cramond, W.R.; Sanders, G.A.; Hatch, S.W.

    1989-04-01

    Shutdown Decay Heat Removal Requirements has been designated as Unresolved Safety Issue (USI) A-45. The overall objectives of the USI A-45 program were to evaluate the safety adequacy of decay heat removal (DHR) systems in existing light water reactor nuclear power plants and to assess the value and impact (benefit-cost) of alternative measures for improving the overall reliability of the DHR function. To provide the technical data required to meet these objectives a program was developed that examined the state of DHR system reliability in a sample of existing plants. This program identified potential vulnerabilities and identified and established the feasibility of potential measures to improve the reliability of the DHR function. A value/impact (V/I) analysis of the more promising of such measures was conducted and documented. This report summarizes those studies. In addition, because of the evolving nature of V/I analyses in support of regulation, a number of supporting studies related to appropriate procedures and measures for the V/I analyses were also conducted. These studies are also summarized herein. This report only summarizes findings of technical studies performed by Sandia National Laboratories as part of the program to resolve this issue. 46 refs., 7 figs., 124 tabs

  18. Removal of Fe(II) from tap water by electrocoagulation technique

    Energy Technology Data Exchange (ETDEWEB)

    Ghosh, D.; Solanki, H. [Department of Chemical Engineering, Indian Institute of Technology, Guwahati 781039 (India); Purkait, M.K. [Department of Chemical Engineering, Indian Institute of Technology, Guwahati 781039 (India)], E-mail: mihir@iitg.ernet.in

    2008-06-30

    Electrocoagulation (EC) is a promising electrochemical technique for water treatment. In this work electrocoagulation (with aluminum as electrodes) was studied for iron Fe(II) removal from aqueous medium. Different concentration of Fe(II) solution in tap water was considered for the experiment. During EC process, various amorphous aluminum hydroxides complexes with high sorption capacity were formed. The removal of Fe(II) was consisted of two principal steps; (a) oxidation of Fe(II) to Fe(III) and (b) subsequent removal of Fe(III) by the freshly formed aluminum hydroxides complexes by adsorption/surface complexation followed by precipitation. Experiments were carried out with different current densities ranging from 0.01 to 0.04 A/m{sup 2}. It was observed that the removal of Fe(II) increases with current densities. Inter electrode distance was varied from 0.005 to 0.02 m and was found that least inter electrode distance is suitable in order to achieve higher Fe(II) removal. Other parameters such as conductivity, pH and salt concentration were kept constant as per tap water quality. Satisfactory iron removal of around 99.2% was obtained at the end of 35 min of operation from the initial concentration of 25 mg/L Fe(II). Iron concentration in the solution was determined using Atomic absorption spectrophotometer. By products obtained from the electrocoagulation bath were analyzed by SEM image and corresponding elemental analysis (EDAX). Cost estimation for the electrocoagulation was adopted and explained well. Up to 15 mg/L of initial Fe(II) concentration, the optimum total cost was 6.05 US$/m{sup 3}. The EC process for removing Fe(II) from tap water is expected to be adaptable for household use.

  19. Investigation on natural convection decay heat removal for the EFR: Status of the program

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H; Weinberg, D [Kernforschungszentrum Karlsruhe GmbH, IATF, Karlsruhe (Germany); Webster, R [AEA Reactor Services, Dounreay (United Kingdom)

    1991-07-01

    The European Research and Development Program on decay heat removal by natural convection for the European Fast Reactor (EFR) covers the calculational methods and the model experiments performed for code validation. The studies concentrate on important physical effects of the cooling modes withinthe primary system and the direct reactor cooling circuits and include fundamental tests as well as reactor experiments. (author)

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

    International Nuclear Information System (INIS)

    Crossley, H.; Wakefield, J.R.

    1991-01-01

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

  1. Unavailability of the residual system heat removal of Angra 1 by Bayesian networks considering dependent failures

    International Nuclear Information System (INIS)

    Gomes, Many R.S.; Melo, Paulo F.F.F. e

    2015-01-01

    This work models by Bayesian networks the residual heat removal system (SRCR) of Angra I nuclear power plant, using fault tree mapping for systematically identifying all possible modes of occurrence caused by a large loss of coolant accident (large LOCA). The focus is on dependent events, such as the bridge system structure of the residual heat removal system and the occurrence of common-cause failures. We used the Netica™ tool kit, Norsys Software Corporation and Python 2.7.5 for modeling Bayesian networks and Microsoft Excel for modeling fault trees. Working with dependent events using Bayesian networks is similar to the solutions proposed by other models, beyond simple understanding and ease of application and modification throughout the analysis. The results obtained for the unavailability of the system were satisfactory, showing that in most cases the system will be available to mitigate the effects of an accident as described above. (author)

  2. Unavailability of the residual system heat removal of Angra 1 by Bayesian networks considering dependent failures

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Many R.S.; Melo, Paulo F.F.F. e, E-mail: mgomes@con.ufrj.br, E-mail: frutuoso@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Programa de Pos-Graduacao em Engenharia Nuclear

    2015-07-01

    This work models by Bayesian networks the residual heat removal system (SRCR) of Angra I nuclear power plant, using fault tree mapping for systematically identifying all possible modes of occurrence caused by a large loss of coolant accident (large LOCA). The focus is on dependent events, such as the bridge system structure of the residual heat removal system and the occurrence of common-cause failures. We used the Netica™ tool kit, Norsys Software Corporation and Python 2.7.5 for modeling Bayesian networks and Microsoft Excel for modeling fault trees. Working with dependent events using Bayesian networks is similar to the solutions proposed by other models, beyond simple understanding and ease of application and modification throughout the analysis. The results obtained for the unavailability of the system were satisfactory, showing that in most cases the system will be available to mitigate the effects of an accident as described above. (author)

  3. Tritium retention in next step devices and the requirements for mitigation and removal techniques

    International Nuclear Information System (INIS)

    Counsell, G; Coad, P; Grisola, C; Hopf, C

    2006-01-01

    Mechanisms underlying the retention of fuel species in tokamaks with carbon plasma-facing components are presented, together with estimates for the corresponding retention of tritium in ITER. The consequential requirement for new and improved schemes to reduce the tritium inventory is highlighted and the results of ongoing studies into a range of techniques are presented, together with estimates of the tritium removal rate in ITER in each case. Finally, an approach involving the integration of many tritium removal techniques into the ITER operational schedule is proposed as a means to extend the period of operations before major intervention is required

  4. Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

  5. Local Heat Transfer and CHF for Subcooled Flow Boiling - Annual Report 1993

    International Nuclear Information System (INIS)

    Boyd, Ronald D.

    2000-01-01

    Subcooled flow boiling in heated coolant channels is an important heat transfer enhancement technique in the development of fusion reactor components, where high heat fluxes must be accommodated. As energy fluxes increase in magnitude, additional emphasis must be devoted to enhancing techniques such as sub cooling and enhanced surfaces. In addition to subcooling, other high heat flux alternatives such as high velocity helium and liquid metal cooling have been considered as serious contenders. Each technique has its advantages and disadvantages [1], which must be weighed as to reliability and reduced cost of fusion reactor components. Previous studies [2] have set the stage for the present work, which will concentrate on fundamental thermal hydraulic issues associated with the h-international Thermonuclear Experimental Reactor (ITER) and the Engineering Design Activity (EDA). This proposed work is intended to increase our understanding of high heat flux removal alternatives as well as our present capabilities by: (1) including single-side heating effects in models for local predictions of heat transfer and critical heat flux; (2) inspection of the US, Japanese, and other possible data sources for single-side heating, with the aim of exploring possible correlations for both CHF and local heat transfer; and (3) assessing the viability of various high heat flux removal techniques. The latter task includes: (a) sub-cooled water flow boiling with enhancements such as twisted tapes, and hypervapotrons, (b) high velocity helium cooling, and (c) other potential techniques such as liquid metal cooling. This assessment will increase our understanding of: (1) hypervapotron heat transfer via fins, flow recirculation, and flow oscillation, and (2) swirl flow. This progress report contains selective examples of ongoing work. Section II contains an extended abstract, which is part of and evolving technical paper on single-side f heating. Section III describes additional details

  6. High heat flux cooling for accelerator targets

    International Nuclear Information System (INIS)

    Silverman, I.; Nagler, A.

    2002-01-01

    Accelerator targets, both for radioisotope production and for high neutron flux sources generate very high thermal power in the target material which absorbs the particles beam. Generally, the geometric size of the targets is very small and the power density is high. The design of these targets requires dealing with very high heat fluxes and very efficient heat removal techniques in order to preserve the integrity of the target. Normal heat fluxes from these targets are in the order of 1 kw/cm 2 and may reach levels of an order of magnitude higher

  7. Studies on the characteristics of the separated heat pipe system with non-condensible gas for the use of the passive decay heat removal in reactor systems

    International Nuclear Information System (INIS)

    Hayashi, Takao; Ishi, Takayuki; Hayakawa, Hitoshi; Ohashi, Kazutaka

    1997-01-01

    Experiments on the separated heat pipe system of variable conductance type, which enclose non-condensible gas, have been carried out with intention of applying such system to passive decay heat removal of the modular reactors such as HTR plant. Basic experiments have been carried out on the experimental apparatus consisting of evaporator, vapor transfer tube, condenser tube and return tube which returns the condensed liquid back to the evaporator. Water and methanol were examined as the working fluids and nitrogen gas was enclosed as the non-condensible gas. The behaviors of the system were examined for the parametric changes of the heat input under the various pressures of nitrogen gas initially enclosed, including the case without enclosing N 2 gas for the comparison. The results of the experiments shows very clear features of self control characteristics. The self control mechanism was made clear, that is, in such system in which the condensing area in the condenser expands automatically in accordance with the increase of the heat input to keep the system temperature nearly constant. The working temperature of the system are clearly dependent on the pressure of the non-condensable gas initially enclosed, with higher system working temperature with higher initial gas pressure enclosed. The analyses were done on water and methanol as the working fluids, which show very good agreement with the experimental results. A lot of attractive applications are expected including the self switching feature with minimum heat loss during normal operation with maintaining the sufficient heat removal at accidents. (author)

  8. Fabrication techniques to eliminate postweld heat treatment

    International Nuclear Information System (INIS)

    Lochhead, J.C.

    1978-01-01

    Postweld heat treatments to reduce residual stresses (stress relief operations) have been a common practice in the pressure vessel industry for a large number of years. A suitable heat treatment operation can, in particular for low alloy steels, have additional beneficial effects, i.e. a reduction in peak hardness values in the heat-affected zone, an improvement in weld metal properties, and a lowering of the adverse effects of the welding process on the mechanical properties of the parent material adjacent to the weld metal. However, continuing studies in the field of brittle fracture, improved parent materials, and more sophisticated nondestructive testing techniques have led to the elimination of such a practice in ever-increasing thickness ranges and types of material. For instance, the recently issued BS 5500 compared with BS 1113 (1969) lifts the thickness limit requiring stress relief in certain circumstances from 19 to 35mm for C steels. With respect to materials the CEGB has stated that as a result of successful operational experience it will no longer be necessary to postweld heat treat butt welds in 2 1/4 Cr-1Mo tubes of certain dimensions. Despite this trend, over a period of years a number of instances have arisen where, because of some factor, postweld heat treatment, although perhaps desirable, is not possible. This Paper describes several such examples. It must be noted that the examples quoted consist of relatively important and major items. It has been necessary within the confines of this Paper to condense the reports. It is hoped that no significant factors have been omitted. (author)

  9. Ultrasound guided percutaneous removal of wooden foreign bodies in the extremities with hydro-dissection technique

    Energy Technology Data Exchange (ETDEWEB)

    Park, HeeJin; Lee, So Yeon; Chung, Eun Chul; Rho, Myung Ho [Dept. of Radiology, Sungkyunkwan University School of Medicine, Kangbuk Samsung Hospital, Seoul (Korea, Republic of); Lee, Sung Moon; Son, Eun Seok [Dongsan Medical Center, Keimyung University School of Medicine, Daegu (Korea, Republic of); Lee, Sun Joo [Dept. of Radiology, Inje University College of Medicine, Busan Paik Hospital, Busan (Korea, Republic of)

    2015-12-15

    We described the technique of ultrasound (US)-guided percutaneous removal of the foreign bodies (FB) with hydro-dissection in the radiologic department and presented video files of several cases.Four patients referred to the radiology department for US evaluation and US-guided percutaneous removal of the FBs in the upper and lower extremities between November, 2006 and November, 2013 were included in this study. The procedures started with US evaluation for the exact location and shape of the FB. A 5 mm-sized skin incision was made at the site of the nearest point from the FB where no passing arteries or tendons were present. We adopted a hydro dissection technique to separate the FB from adjacent tissue using a 2% lidocaine solution. Injected anesthetics detached the FBs from surrounding tissue and thereby facilitated removal. After the tip of the mosquito forceps reached the FB, the wooden FBs were removed. The mean time required for the entire procedure was approximately 20 minutes. There were no significant complications during the US-guided removal or long-term complications after the procedure. All 4 FBs were successfully removed from the soft tissue under US guidance. Ultrasound-guided percutaneous removal of the FBs with hydro-dissection in the radiology department is a less invasive and safe method over surgical removal in the operating room. Additionally, the use of a guide wire and serial dilator may help minimize soft tissue injury and facilitate the introduction of forceps.

  10. Reliability evaluation of power supply and distribution for special heat removal systems in nuclear power stations

    International Nuclear Information System (INIS)

    Jazbec, D.

    1982-01-01

    An example of the power supply and distribution of a Special Emergency Heat Removal System (SEHR) shows how an engineering organization may, with the aid of the analytical method of min-cut sets optimize the system reliability. Herein are given the necessary simple calculation methods. (Auth.)

  11. A review of computer-aided design/computer-aided manufacture techniques for removable denture fabrication

    Science.gov (United States)

    Bilgin, Mehmet Selim; Baytaroğlu, Ebru Nur; Erdem, Ali; Dilber, Erhan

    2016-01-01

    The aim of this review was to investigate usage of computer-aided design/computer-aided manufacture (CAD/CAM) such as milling and rapid prototyping (RP) technologies for removable denture fabrication. An electronic search was conducted in the PubMed/MEDLINE, ScienceDirect, Google Scholar, and Web of Science databases. Databases were searched from 1987 to 2014. The search was performed using a variety of keywords including CAD/CAM, complete/partial dentures, RP, rapid manufacturing, digitally designed, milled, computerized, and machined. The identified developments (in chronological order), techniques, advantages, and disadvantages of CAD/CAM and RP for removable denture fabrication are summarized. Using a variety of keywords and aiming to find the topic, 78 publications were initially searched. For the main topic, the abstract of these 78 articles were scanned, and 52 publications were selected for reading in detail. Full-text of these articles was gained and searched in detail. Totally, 40 articles that discussed the techniques, advantages, and disadvantages of CAD/CAM and RP for removable denture fabrication and the articles were incorporated in this review. Totally, 16 of the papers summarized in the table. Following review of all relevant publications, it can be concluded that current innovations and technological developments of CAD/CAM and RP allow the digitally planning and manufacturing of removable dentures from start to finish. As a result according to the literature review CAD/CAM techniques and supportive maxillomandibular relationship transfer devices are growing fast. In the close future, fabricating removable dentures will become medical informatics instead of needing a technical staff and procedures. However the methods have several limitations for now. PMID:27095912

  12. Application of the PSA method to decay heat removal systems in a large scale FBR design

    International Nuclear Information System (INIS)

    Kotake, S.; Satoh, K.; Matsumoto, H.; Sugawara, M.; Sakata, K.; Okabe, A.

    1993-01-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10 -7 /d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  13. Application of the PSA method to decay heat removal systems in a large scale FBR design

    Energy Technology Data Exchange (ETDEWEB)

    Kotake, S; Satoh, K [Japan Atomic Power Company, Otemachi, Chiyoda-ku, Tokyo (Japan); Matsumoto, H; Sugawara, M [Toshiba Corporation (Japan); Sakata, K [Mitsubishi Atomic Power Industries Inc. (Japan); Okabe, A [Hitachi Engineering Co., Ltd. (Japan)

    1993-02-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10{sup -7}/d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  14. The steady-state modeling and optimization of a refrigeration system for high heat flux removal

    International Nuclear Information System (INIS)

    Zhou Rongliang; Zhang Tiejun; Catano, Juan; Wen, John T.; Michna, Gregory J.; Peles, Yoav; Jensen, Michael K.

    2010-01-01

    Steady-state modeling and optimization of a refrigeration system for high heat flux removal, such as electronics cooling, is studied. The refrigeration cycle proposed consists of multiple evaporators, liquid accumulator, compressor, condenser and expansion valves. To obtain more efficient heat transfer and higher critical heat flux (CHF), the evaporators operate with two-phase flow only. This unique operating condition necessitates the inclusion of a liquid accumulator with integrated heater for the safe operation of the compressor. Due to the projected incorporation of microchannels into the system to enhance the heat transfer in heat sinks, the momentum balance equation, rarely seen in previous vapor compression cycle heat exchangers modeling efforts, is utilized in addition to the mass and energy balance equations to capture the expected significant microchannel pressure drop witnessed in previous experimental investigations. Using the steady-state model developed, a parametric study is performed to study the effect of various external inputs on the system performance. The Pareto optimization is applied to find the optimal system operating conditions for given heat loads such that the system coefficient of performance (COP) is optimized while satisfying the CHF and other system operation constraints. Initial validation efforts show the good agreement between the experimental data and model predictions.

  15. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  16. Summary report of NEPTUN investigations into the steady state thermal hydraulics of the passive decay heat removal

    International Nuclear Information System (INIS)

    Rust, K.; Weinberg, D.; Hoffmann, H.; Frey, H.H.; Baumann, W.; Hain, K.; Leiling, W.; Hayafune, H.; Ohira, H.

    1995-12-01

    During the course of steady state NEPTUN investigations, the effects of different design and operating parameters were studied; in particular: The shell design of the above core sturcture, the core power, the number of decay heat exchangers put in operation, the complete flow path blockage at the primary side of the intermediate heat exchangers, and the fluid level in the primary vessel. The findings of the NEPTUN experiments indicate that the decay heat can be safely removed by natural convection. The interwrapper flow makes an essential contribution to that behavior. The decay heat exchangers installed in the upper plenum cause a thermal stratification associated with a pronounced gradient. The vertical extent of the stratification and the quantity of the gradient are depending on the fact whether a permeable or an impermeable shell covers the above core structure. An increase of the core power or a reduction of the number of decay heat exchangers being in operation leads to a higher temperature level in the primary system but does not alter the global temperature distribution. In the case that no coolant enters the inlet windows at the primary side of the intermediate and decay heat exchangers, the core remains coolable as far as the primary vessel is filled with fluid up to a minimum level. Cold water penetrates from the upper plenum into the core and removes the decay heat. The thermal hydraulic computer code FLUTAN was applied for the three-dimensional numerical simulation of the majority of NEPTUN tests reported here. The comparison of computed against experimental data indicates a qualitatively and quantitatively satisfying agreement of the findings with respect to the field of isotherms as well as the temperature profiles in the upper plenum and within the core region of very complex geometry. (orig./HP) [de

  17. Novel localized heating technique on centrifugal microfluidic disc with wireless temperature monitoring system.

    Science.gov (United States)

    Joseph, Karunan; Ibrahim, Fatimah; Cho, Jongman

    2015-01-01

    Recent advances in the field of centrifugal microfluidic disc suggest the need for electrical interface in the disc to perform active biomedical assays. In this paper, we have demonstrated an active application powered by the energy harvested from the rotation of the centrifugal microfluidic disc. A novel integration of power harvester disc onto centrifugal microfluidic disc to perform localized heating technique is the main idea of our paper. The power harvester disc utilizing electromagnetic induction mechanism generates electrical energy from the rotation of the disc. This contributes to the heat generation by the embedded heater on the localized heating disc. The main characteristic observed in our experiment is the heating pattern in relative to the rotation of the disc. The heating pattern is monitored wirelessly with a digital temperature sensing system also embedded on the disc. Maximum temperature achieved is 82 °C at rotational speed of 2000 RPM. The technique proves to be effective for continuous heating without the need to stop the centrifugal motion of the disc.

  18. Study of Modern Nano Enhanced Techniques for Removal of Dyes and Metals

    Directory of Open Access Journals (Sweden)

    Samavia Batool

    2014-01-01

    Full Text Available Industrial effluent often contains the significant amount of hexavalent chromium and synthetic dyes. The discharge of wastewater without proper treatment into water streams consequently enters the soil and disturbs the aquatic and terrestrial life. A range of wastewater treatment technologies have been proposed which can efficiently reduce both Cr(VI and azo dyes simultaneously to less toxic form such as biodegradation, biosorption, adsorption, bioaccumulation, and nanotechnology. Rate of simultaneous reduction of Cr(VI and azo dyes can be enhanced by combining different treatment techniques. Utilization of synergistic treatment is receiving much attention due to its enhanced efficiency to remove Cr(VI and azo dye simultaneously. This review evaluates the removal methods for simultaneous removal of Cr(VI and azo dyes by nanomicrobiology, surface engineered nanoparticles, and nanophotocatalyst. Sorption mechanism of biochar for heavy metals and organic contaminants is also discussed. Potential microbial strains capable of simultaneous removal of Cr(VI and azo dyes have been summarized in some details as well.

  19. Sabots, Obturator and Gas-In-Launch Tube Techniques for Heat Flux Models in Ballistic Ranges

    Science.gov (United States)

    Bogdanoff, David W.; Wilder, Michael C.

    2013-01-01

    For thermal protection system (heat shield) design for space vehicle entry into earth and other planetary atmospheres, it is essential to know the augmentation of the heat flux due to vehicle surface roughness. At the NASA Ames Hypervelocity Free Flight Aerodynamic Facility (HFFAF) ballistic range, a campaign of heat flux studies on rough models, using infrared camera techniques, has been initiated. Several phenomena can interfere with obtaining good heat flux data when using this measuring technique. These include leakage of the hot drive gas in the gun barrel through joints in the sabot (model carrier) to create spurious thermal imprints on the model forebody, deposition of sabot material on the model forebody, thereby changing the thermal properties of the model surface and unknown in-barrel heating of the model. This report presents developments in launch techniques to greatly reduce or eliminate these problems. The techniques include the use of obturator cups behind the launch package, enclosed versus open front sabot designs and the use of hydrogen gas in the launch tube. Attention also had to be paid to the problem of the obturator drafting behind the model and impacting the model. Of the techniques presented, the obturator cups and hydrogen in the launch tube were successful when properly implemented

  20. Passive decay heat removal by natural air convection after severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Erbacher, F.J.; Neitzel, H.J. [Forschungszentrum Karlsruhe Institut fur Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany); Cheng, X. [Technische Universitaet Karlsruhe Institut fur Stroemungslehre und Stroemungsmaschinen, Karlsruhe (Germany)

    1995-09-01

    The composite containment proposed by the Research Center Karlsruhe and the Technical University Karlsruhe is to cope with severe accidents. It pursues the goal to restrict the consequences of core meltdown accidents to the reactor plant. One essential of this new containment concept is its potential to remove the decay heat by natural air convection and thermal radiation in a passive way. To investigate the coolability of such a passive cooling system and the physical phenomena involved, experimental investigations are carried out at the PASCO test facility. Additionally, numerical calculations are performed by using different codes. A satisfying agreement between experimental data and numerical results is obtained.

  1. Time evolution simulation of heat removal in a small water tank by natural convection

    Energy Technology Data Exchange (ETDEWEB)

    Freitas, Carlos Alberto de, E-mail: carlos.freitas1950@hotmail.com [Instituto Federal do Rio de Janeiro (IFRJ), Nilopolis, RJ (Brazil); Jachic, Joao; Moreira, Maria de Lourdes, E-mail: jjachic@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    One of the cooling modes for any source of heat such as in a shutdown nuclear core is the natural convection. The design specifications of any cooling pool can only be done when the removal heat rate and the corresponding mass flow rate is reasonably established. In our simulation scheme, we assumed that the body forces acting in the cubic water cell are: the weight, the drag force and the integrated pressure forces on the horizontal surfaces, the viscosity shear forces on the vertical surfaces and also a special viscosity drag force due to the mass dislocation along a Bernoulli type current tube outside the motive region. For a suitable time step, the uprising convection velocity is determined by an implicit and also by an explicit solution algorithm. The resulting differential equation depends on updating specific mass, dynamic viscosity and constant pressure heat coefficient with the last known temperature in the cell that absorbed heat. Numerical calculation software was performed using MATLAB’s technical computing language and then applied for a heat generation plate simulating a spent fuel assembler from a shutdown nuclear core. The results show time evolution of convection, terminal velocity and water temperature distribution. Pool dimension as well as pool level decrement are also determined for various air exhausting system conditions and heat rate of the spent fuel plate being cooled. (author)

  2. Time evolution simulation of heat removal in a small water tank by natural convection

    International Nuclear Information System (INIS)

    Freitas, Carlos Alberto de; Jachic, Joao; Moreira, Maria de Lourdes

    2013-01-01

    One of the cooling modes for any source of heat such as in a shutdown nuclear core is the natural convection. The design specifications of any cooling pool can only be done when the removal heat rate and the corresponding mass flow rate is reasonably established. In our simulation scheme, we assumed that the body forces acting in the cubic water cell are: the weight, the drag force and the integrated pressure forces on the horizontal surfaces, the viscosity shear forces on the vertical surfaces and also a special viscosity drag force due to the mass dislocation along a Bernoulli type current tube outside the motive region. For a suitable time step, the uprising convection velocity is determined by an implicit and also by an explicit solution algorithm. The resulting differential equation depends on updating specific mass, dynamic viscosity and constant pressure heat coefficient with the last known temperature in the cell that absorbed heat. Numerical calculation software was performed using MATLAB’s technical computing language and then applied for a heat generation plate simulating a spent fuel assembler from a shutdown nuclear core. The results show time evolution of convection, terminal velocity and water temperature distribution. Pool dimension as well as pool level decrement are also determined for various air exhausting system conditions and heat rate of the spent fuel plate being cooled. (author)

  3. Evaluation of the dual staining technique for complete removal of idiopathic epimacular membranes

    Directory of Open Access Journals (Sweden)

    Ahmed M Abdel Hadi

    2015-01-01

    In conclusion, dual staining starting with TA effectively led to adequate visualization and removal of the posterior hyaloid as well as the fibrous component of the idiopathic EMM. The subsequent BBG administration in an air filled vitreous (the dry technique helped selective removal of the ILM with no deleterious effects on the functional or the anatomical outcomes of the procedure as demonstrated by the significant improvement in both the BCVA and the CMT 6 months after surgery.

  4. Removable partial denture alloys processed by laser-sintering technique.

    Science.gov (United States)

    Alageel, Omar; Abdallah, Mohamed-Nur; Alsheghri, Ammar; Song, Jun; Caron, Eric; Tamimi, Faleh

    2018-04-01

    Removable partial dentures (RPDs) are traditionally made using a casting technique. New additive manufacturing processes based on laser sintering has been developed for quick fabrication of RPDs metal frameworks at low cost. The objective of this study was to characterize the mechanical, physical, and biocompatibility properties of RPD cobalt-chromium (Co-Cr) alloys produced by two laser-sintering systems and compare them to those prepared using traditional casting methods. The laser-sintered Co-Cr alloys were processed by the selective laser-sintering method (SLS) and the direct metal laser-sintering (DMLS) method using the Phenix system (L-1) and EOS system (L-2), respectively. L-1 and L-2 techniques were 8 and 3.5 times more precise than the casting (CC) technique (p laser-sintered and cast alloys were biocompatible. In conclusion, laser-sintered alloys are more precise and present better mechanical and fatigue properties than cast alloys for RPDs. © 2017 Wiley Periodicals, Inc. J Biomed Mater Res Part B: Appl Biomater, 106B: 1174-1185, 2018. © 2017 Wiley Periodicals, Inc.

  5. Biological and Physiochemical Techniques for the Removal of Zinc from Drinking Water: A Review

    Directory of Open Access Journals (Sweden)

    Naseem Zahra

    2015-12-01

    Full Text Available Presence of Zinc (II in drinking water beyond permissible limits is considered unsafe for human health. Many different anthropogenic activities including mining, burning of petroleum, industrialization, and urbanization cause a release of considerably higher amounts of zinc into the waterbodies. A permissible limit of 5 mg/L is set by various environmental and pollution control authorities beyond which water may cause respiratory, liver, gonads, and brain disorders. Due to these health hazards, it is important to remove exceeding amounts of zinc from drinking water. Zinc enters drinking water from various sources such as corrosive pipelines, release of industrial effluents, and metal leaching. Different biological and physiochemical techniques are used to remove zinc involving chemical precipitation, ion exchange, adsorption, biosorbents, distillation, ozonation, and membrane filtration technology. Among these technologies, physical process of adsorption using low cost adsorbents is not only economical but abundant, efficient, and easily available. In present review different physiochemical and biological techniques are discussed for the removal of Zinc from drinking water.

  6. A Novel Technique Using a Protection Filter During Fibrin Sheath Removal for Implanted Venous Access Device Dysfunction

    Energy Technology Data Exchange (ETDEWEB)

    Sotiriadis, Charalampos; Hajdu, Steven David [University Hospital of Lausanne, Cardiothoracic and Vascular Unit, Department of Radiology (Switzerland); Degrauwe, Sophie [University Hospital of Lausanne, Department of Cardiology (Switzerland); Barras, Heloise; Qanadli, Salah Dine, E-mail: salah.qanadli@chuv.ch [University Hospital of Lausanne, Cardiothoracic and Vascular Unit, Department of Radiology (Switzerland)

    2016-08-15

    With the increased use of implanted venous access devices (IVADs) for continuous long-term venous access, several techniques such as percutaneous endovascular fibrin sheath removal, have been described, to maintain catheter function. Most standard techniques do not capture the stripped fibrin sheath, which is subsequently released in the pulmonary circulation and may lead to symptomatic pulmonary embolism. The presented case describes an endovascular technique which includes stripping, capture, and removal of fibrin sheath using a novel filter device. A 64-year-old woman presented with IVAD dysfunction. Stripping was performed using a co-axial snare to the filter to capture the fibrin sheath. The captured fragment was subsequently removed for visual and pathological verification. No immediate complication was observed and the patient was discharged the day of the procedure.

  7. Terbuthylazine and desethylterbuthylazine: Recent occurrence, mobility and removal techniques.

    Science.gov (United States)

    Tasca, Andrea Luca; Puccini, Monica; Fletcher, Ashleigh

    2018-07-01

    The herbicide terbuthylazine (TBA) has displaced atrazine in most of EU countries, becoming one of the most regularly used pesticides and, therefore, frequently detected in natural waters. The affinity of TBA for soil organic matter suggests prolonged contamination; degradation leads to the release of the metabolite desethylterbuthylazine (DET), which has higher water solubility and binds more weakly to organic matter compared to the parent compound, resulting in higher associated risk for contamination of groundwater resources. Additionally, TBA and DET are chemicals of emerging concern because of their persistence and toxicity towards aquatic organisms; moreover, they are known to have significant endocrine disruption capacity to wildlife and humans. Conventional treatments applied during drinking water production do not lead to the complete removal of these chemicals; activated carbon provides the greatest efficiency, whereas ozonation can generate by-products with comparable oestrogenic activity to atrazine. Hydrogen peroxide alone is ineffective to degrade TBA, while UV/H 2 O 2 advanced oxidation and photocatalysis are the most effective processes for oxidation of TBA. It has been determined that direct photolysis gives the highest degradation efficiency of all UV/H 2 O 2 treatments, while most of the photocatalytic degradation is attributed to OH radicals, and TiO 2 solar-photocatalytic ozonation can lead to almost complete TBA removal in ∼30 min. Constructed wetlands provide a valuable buffer capacity, protecting downstream surface waters from contaminated runoff. TBA and DET occurrence are summarized and removal techniques are critically evaluated and compared, to provide the reader with a comprehensive guide to state-of-the-art TBA removal and potential future treatments. Copyright © 2018 Elsevier Ltd. All rights reserved.

  8. Summary report for Group X6: Heat removal system and system analysis

    Energy Technology Data Exchange (ETDEWEB)

    Leung, W

    2005-12-15

    This report is a summary of the activities of the X6 design support for the Heat Removal System (HRS) of MEGAPIE. It can be divided into two main parts: The first part is about the design and manufacturing of he cooling loop (the first 3 chapters), and the second part is dealing with the thermal hydraulic analysis of the overall HRS. This also reflects the change of the X6 activities from design to operation support. The activities of this group are more or less driven by the needs rather than a complete set of tasks given at the start of the project. The first part chronicles the system development. Some of the arguments are probably outdated but are kept in the original form to illustrate the evolution of concepts. The main objective is, of course, to design a heat removal system that can cool the liquid metal spallation target for a 1 MW proton beam i.e. 1.74 mA in 575 MeV). It is also reckoned that the liquid metal, BE (lead-bismuth-eutectic), must be kept liquid even when the proton beam was switched off. This requires either that the cooling system can be shut down or the operating temperature of the coolant be higher than the freezing point of LBE. As for safety concerns, the HRS system must not exert a pressure that exceeds the design pressure of the target beam window in case of a break at the target heat exchanger (THX); this limits the cover gas pressure to about 4 bar(a). These are the basic design principles that carry through the conceptual and engineering design of he system. The organic coolant Diphyl THT was then chosen, because of its wide range of operating temperature (i.e. from 0 to 340 degC) and high boiling point, and a proven record in industrial applications. (author)

  9. Summary report for Group X6: Heat removal system and system analysis

    International Nuclear Information System (INIS)

    Leung, W.

    2005-12-01

    This report is a summary of the activities of the X6 design support for the Heat Removal System (HRS) of MEGAPIE. It can be divided into two main parts: The first part is about the design and manufacturing of he cooling loop (the first 3 chapters), and the second part is dealing with the thermal hydraulic analysis of the overall HRS. This also reflects the change of the X6 activities from design to operation support. The activities of this group are more or less driven by the needs rather than a complete set of tasks given at the start of the project. The first part chronicles the system development. Some of the arguments are probably outdated but are kept in the original form to illustrate the evolution of concepts. The main objective is, of course, to design a heat removal system that can cool the liquid metal spallation target for a 1 MW proton beam i.e. 1.74 mA in 575 MeV). It is also reckoned that the liquid metal, BE (lead-bismuth-eutectic), must be kept liquid even when the proton beam was switched off. This requires either that the cooling system can be shut down or the operating temperature of the coolant be higher than the freezing point of LBE. As for safety concerns, the HRS system must not exert a pressure that exceeds the design pressure of the target beam window in case of a break at the target heat exchanger (THX); this limits the cover gas pressure to about 4 bar(a). These are the basic design principles that carry through the conceptual and engineering design of he system. The organic coolant Diphyl THT was then chosen, because of its wide range of operating temperature (i.e. from 0 to 340 degC) and high boiling point, and a proven record in industrial applications. (author)

  10. Non-linear effects in vortex viscous flow in superconductors-role of finite heat removal velocity

    International Nuclear Information System (INIS)

    Bezuglyj, A.I.; Shklovskij, V.A.

    1991-01-01

    The role of finite heat removal velocity in experiments on non-linear effects in vortex viscous flow in superconducting films near critical temperature was investigated. It was shown that the account of thermal effects permits to explain the experimentally observed dependence of electron energy relaxation time and current break-down in voltage-current characteristic from magnetic field value. 5 refs.; 1 fig. (author)

  11. Technical specification improvements to containment heat removal and emergency core cooling systems: Final report

    International Nuclear Information System (INIS)

    Sullivan, W.P.; Ha, C.; Pentzien, D.C.; Visweswaran, S.

    1988-07-01

    This report presents the results of an analysis for technical specification improvements to the emergency core cooling systems (ECCS) and containment heat removal systems (EPRI Research Project 2142-3). The objective of this project is to further develop a reliability- and risk-based methodology to provide improvements by considering groups of surveillance test intervals and allowed out-of-service times jointly. This was done for the technical specifications for the ECCS, containment heat removal equipment, and supporting systems of a boiling water reactor plant. The project (1) developed a methodology for optimizing groups of surveillance test intervals and allowed out-of-service times jointly, (2) applied the methodology in a case study of a specific operating plant, Hatch-2, and (3) evaluated benefits of the application. The results of the case study demonstrate that beneficial technical specification improvements can be realized with application of the methodology. By tightening a small group of sensitive surveillance test intervals (STIs) and allowed out-of-service times (AOTs), a larger group of less sensitive STIs and AOTs can be extended resulting in an overall plant operating cost improvement without reducing the plant safety. The reliability- and risk-based methodology and results from this project can be effectively applied for technical specification improvements at other operating plants

  12. Gas removal technique to maintain global environment. Chikyu kankyo hozen no tame no bojo gijutsu

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, K [The University of Tokyo, Tokyo (Japan). Faculty of Engineering

    1992-10-12

    This paper describes the removal technique of gases such as CO2, SO2 and NOx which have the deep relation to the maintenance of global environment. This paper describes partially r SO2 and NOx which are the primary cause of acid rain. As for the removal of CO2 generated from fixed sources (thermal power stations and others), the separation technique and isolation-fixation technique have been researched on and developed. Of the separation method, the effect of the chemical absorption method and the adsorption method is proved with the preceding experiments. The isolation method is differently researched on as to store under deep sea or ground but may be urgent and temporary. The fixation of CO2 is a serious global problem which relates to the afforestation and forests. The fixation which uses coral reefs in ocean as the absorption source has a potential. As for the processing of substances causing acid rain, the desulfurization from petroleum and the flue gas desulfurization have the excellent results. The improvement of combustion method or the flue gas denitrification at the fixed sources are used to remove NOx. The removal of NOx from all diesel cars is difficult compared with the exhaust gas cleaning of gasoline cars and is not commercialized. 11 refs., 1 fig., 2 tabs.

  13. Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

    2003-01-01

    The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

  14. The status of thermal-hydraulic studies on the decay heat removal by natural convection using RAMONA and NEPTUN models

    International Nuclear Information System (INIS)

    Hoffmann, H.; Hain, K.; Marten, K.; Rust, K.; Weinberg, D.; Ohira, H.

    2004-01-01

    Thermal-hydraulic experiments were performed with water in order to simulate the decay heat removal by natural convection in a pool-type sodium-cooled reactor. Two test rigs of different scales were used, namely RAMONA (1:20) and NEPTUN (1:5). RAMONA served to study the transition from nominal operation by forced convection to decay heat removal operation by natural convection. Steady-state similarity tests were carried out in both facilities. The investigations cover nominal and non-nominal operation conditions. These data provide a broad basis for the verification of computer programs. Numerical analyses performed with the three-dimensional FLUTAN code indicated that the thermal-hydraulic processes can be quantitatively simulated even for the very complex geometry of the NEPTUN test rig. (author)

  15. How Effective are Existing Arsenic Removal Techniques

    Science.gov (United States)

    This presentation will summarize the system performance results of the technologies demonstrated in the arsenic demonstration program. The technologies include adsorptive media, iron removal, iron removal with iron additions, iron removal followed by adsorptive media, coagulatio...

  16. Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yavuz, U. [Turkish Atomic Energy Authority, Ankara (Turkey). Nuclear Safety Dept.; Zabunoolu, O.H. [Hacettepe Univ., Ankara (Turkey). Dept. of Nuclear Engineering

    2006-08-15

    Spent nuclear fuel resulting from reactor operation must be safely stored and managed prior to reprocessing and/or final disposal of high-level waste. Any spent fuel storage system must provide for safe receipt, handling, retrieval, and storage of spent fuel. In order to achieve the safe storage, the design should primarily provide for radiation protection, subcriticality of spent fuel, and removal of spent fuel residual heat. This article is focused on the design of a metal-shielded dry-cask storage system, which will host spent LWR fuels burned to 33 000, 45 000, and 55 000 MWd/t U and cooled for 5 or 10 years after discharge from reactor. The storage system is analyzed by taking into account radiation protection, subcriticality, and heat-removal aspects; and appropriate designs, in accordance with the international standards. (orig.)

  17. A titration model for evaluating calcium hydroxide removal techniques

    Directory of Open Access Journals (Sweden)

    Mark PHILLIPS

    2015-02-01

    Full Text Available Objective Calcium hydroxide (Ca(OH2 has been used in endodontics as an intracanal medicament due to its antimicrobial effects and its ability to inactivate bacterial endotoxin. The inability to totally remove this intracanal medicament from the root canal system, however, may interfere with the setting of eugenol-based sealers or inhibit bonding of resin to dentin, thus presenting clinical challenges with endodontic treatment. This study used a chemical titration method to measure residual Ca(OH2 left after different endodontic irrigation methods. Material and Methods Eighty-six human canine roots were prepared for obturation. Thirty teeth were filled with known but different amounts of Ca(OH2 for 7 days, which were dissolved out and titrated to quantitate the residual Ca(OH2 recovered from each root to produce a standard curve. Forty-eight of the remaining teeth were filled with equal amounts of Ca(OH2 followed by gross Ca(OH2 removal using hand files and randomized treatment of either: 1 Syringe irrigation; 2 Syringe irrigation with use of an apical file; 3 Syringe irrigation with added 30 s of passive ultrasonic irrigation (PUI, or 4 Syringe irrigation with apical file and PUI (n=12/group. Residual Ca(OH2 was dissolved with glycerin and titrated to measure residual Ca(OH2 left in the root. Results No method completely removed all residual Ca(OH2. The addition of 30 s PUI with or without apical file use removed Ca(OH2 significantly better than irrigation alone. Conclusions This technique allowed quantification of residual Ca(OH2. The use of PUI (with or without apical file resulted in significantly lower Ca(OH2 residue compared to irrigation alone.

  18. Optimization of residual heat removal pump axial thrust and axial bearing

    International Nuclear Information System (INIS)

    Schubert, F.

    1996-01-01

    The residual heat removal (RHR) pumps of German 1300 megawatt pressurized-water reactor (PWR) power plants are of the single stage end suction type with volute casing or with diffuser and forged circular casing. Due to the service conditions the pumps have to cover the full capacity range as well as a big variation in suction static pressure. This results in a big difference in the axial thrust that has to be borne by the axial bearing. Because these pumps are designed to operate without auxiliary systems (things that do not exist can not fail), they are equipped with antifriction bearings and sump oil lubrication. To minimize the heat production within the bearing casing, a number of PWR plants have pumps with combined axial/radial bearings of the ball type. Due to the fact that the maximum axial thrust caused by static pressure and hydrodynamic forces on the impeller is too big to be borne by that type of axial bearing, the impellers were designed to produce a hydrodynamic axial force that counteracts the static axial force. Thus, the resulting axial thrust may change direction when the static pressure varies

  19. Optimization of residual heat removal pump axial thrust and axial bearing

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, F.

    1996-12-01

    The residual heat removal (RHR) pumps of German 1300 megawatt pressurized-water reactor (PWR) power plants are of the single stage end suction type with volute casing or with diffuser and forged circular casing. Due to the service conditions the pumps have to cover the full capacity range as well as a big variation in suction static pressure. This results in a big difference in the axial thrust that has to be borne by the axial bearing. Because these pumps are designed to operate without auxiliary systems (things that do not exist can not fail), they are equipped with antifriction bearings and sump oil lubrication. To minimize the heat production within the bearing casing, a number of PWR plants have pumps with combined axial/radial bearings of the ball type. Due to the fact that the maximum axial thrust caused by static pressure and hydrodynamic forces on the impeller is too big to be borne by that type of axial bearing, the impellers were designed to produce a hydrodynamic axial force that counteracts the static axial force. Thus, the resulting axial thrust may change direction when the static pressure varies.

  20. Development of core hot spot evaluation method for decay heat removal by natural circulation under transient conditions in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu; Ohkubo, Yoshiyuki

    2010-01-01

    Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot under transient conditions from forced to natural circulation operations that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal, e.g., flow redistribution in fuel assemblies caused by buoyancy force, and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional 1D and detailed 3D simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness. (author)

  1. Removal of a Dental Implant Displaced into the Maxillary Sinus by Means of the Bone Lid Technique

    Directory of Open Access Journals (Sweden)

    Pietro Fusari

    2013-01-01

    Full Text Available Background. Rehabilitation of edentulous jaws with implant-supported prosthesis has become a common practice among oral surgeons in the last three decades. This therapy presents a very low incidence of complications. One of them is the displacement of dental implants into the maxillary sinus. Dental implants, such as any other foreign body into the maxillary sinus, should be removed in order to prevent sinusitis. Methods. In this paper, we report a case of dental implant migrated in the maxillary sinus and removed by means of the bone lid technique. Results and Conclusion. The migration of dental implants into the maxillary sinus is rarely reported. Migrated implants should be considered for removal in order to prevent possible sinusal diseases. The implant has been removed without any complications, confirming the bone lid technique to be safe and reliable.

  2. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  3. Passive decay heat removal by sump cooling after core meltdown

    International Nuclear Information System (INIS)

    Knebel, J.U.; Mueller, U.

    1996-01-01

    This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives first measurement results of the 1:20 linearly scaled plane two-dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototype conditions

  4. An experimental investigation of heat transfer enhancement in minichannel: Combination of nanofluid and micro fin structure techniques

    DEFF Research Database (Denmark)

    Zhang, Ji; Diao, Yanhua; Zhao, Yaohua

    2017-01-01

    This work experimentally studied the single-phase heat transfer and pressure drop characteristics by using two heat transfer enhancement techniques (micro fin structure and nanofluids) in multiport minichannel flat tube (MMFT). MMFT consisted of numerous parallel rectangular minichannels...... and is widely used in industry as the heat transfer unit of a heat exchanger. Firstly, the enhanced heat transfer performances by individually using one enhancement technique were investigated by testing Nusselt number, friction factor and performance evaluation criterion (PEC). In this section, five MMFTs...... with different micro fin numbers (N = 0, 1, 2, 3 and 4) and nanofluids with three volume concentrations (φ = 0.005%, 0.01% and 0.1%) were used as test sections and working fluids respectively. Secondly, the experiments using two combined enhancement technique were performed. By using conjunctively two...

  5. Final-impression techniques and materials for making complete and removable partial dentures.

    Science.gov (United States)

    Jayaraman, Srinivasan; Singh, Balendra P; Ramanathan, Balasubramanian; Pazhaniappan Pillai, Murukan; MacDonald, Laura; Kirubakaran, Richard

    2018-04-04

    Endentulism is relatively common and is often treated with the provision of complete or partial removable dentures. Clinicians make final impressions of complete dentures (CD) and removable partial dentures (RPD) using different techniques and materials. Applying the correct impression technique and material, based on an individual's oral condition, improves the quality of the prosthesis, which may improve quality of life. To assess the effects of different final-impression techniques and materials used to make complete dentures, for retention, stability, comfort, and quality of life in completely edentulous people.To assess the effects of different final-impression techniques and materials used to make removable partial dentures, for stability, comfort, overextension, and quality of life in partially edentulous people. Cochrane Oral Health's Information Specialist searched the following databases: Cochrane Oral Health's Trials Register (to 22 November 2017), the Cochrane Central Register of Controlled Trials (CENTRAL) (Cochrane Register of Studies, to 22 November 2017), MEDLINE Ovid (1946 to 22 November 2017), and Embase Ovid (21 December 2015 to 22 November 2017). The US National Institutes of Health Trials Registry (ClinicalTrials.gov) and the World Health Organization International Clinical Trials Registry Platform were searched for ongoing trials. No restrictions were placed on language or publication status when searching the electronic databases, however the search of Embase was restricted by date due to the Cochrane Centralised Search Project to identify all clinical trials and add them to CENTRAL. We included randomised controlled trials (RCTs) comparing different final-impression techniques and materials for treating people with complete dentures (CD) and removable partial dentures (RPD). For CD, we included trials that compared different materials or different techniques or both. In RPD for tooth-supported conditions, we included trials comparing the

  6. Fish hook injury: removal by ''push through and cut off'' technique: a case report and brief literature review.

    Science.gov (United States)

    Ahmad Khan, Hayat; Kamal, Younis; Lone, Ansar Ul Haq

    2014-04-01

    Fishing is a leisure activity for some people around the world. Accidently the fish hook can get hooked in the hand. If the hook is barbed, removal becomes difficult. We report a case of such a injury in the hand and discuss the technique for its removal with a brief review of the literature. A thirty-two year old male accidently suffered a fishhook injury to his hand. He came to the orthopaedic ward two hours after the incident with pain; the fish hook was hanging from the hand. Unsuccessful attempts to remove it were made by his relatives. A push-through and cut-off technique was used for removal of barbed hook. Barbed hooks are to be removed atraumatically with controlled incision over properly anaesthetised skin. Proper wound management and prophylactic antibiotics suitable for treatment of Aeromonas species should be initiated to prevent complications.

  7. Features of an emergency heat-conducting path in reactors about lead-bismuth and lead heat-carriers

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Bokova, T.A.; Molodtsov, A.A.

    2006-01-01

    The reactor emergency heat removal systems should transfer heat from the surface of reactor core fuel element claddings to the primary circuit followed by heat transfer to the environment. One suggests three design approaches for emergency heat removal systems in lead-bismuth and lead cooled reactor circuits that take account of the peculiar nature of their features. Application of the discussed systems for emergency heat removal improves safety of lead-bismuth and lead cooled reactor plants [ru

  8. Loss of residual heat removal system: Diablo Canyon, Unit 2, April 10, 1987

    International Nuclear Information System (INIS)

    1987-06-01

    This report presents the findings of an NRC Augmented Inspection Team (AIT) investigation into the circumstances associated with the loss of residual heat removal (RHR) system capability for a period of approximately one and one-half hours at the Diablo Canyon, Unit 2 reactor facility on April 10, 1987. This event occurred while the Diablo Canyon, Unit 2, a pressurized water reactor, was shutdown with the reactor coolant system (RCS) water level drained to approximately mid-level of the hot leg piping. The reactor containment building equipment hatch was removed at the time of the event, and plant personnel were in the process of removing the primary side manways to gain access into the steam generator channel head areas. Thus, two fission product barriers were breached throughout the event. The RCS temperature increased from approximately 87 0 F to bulk boiling conditions without RCS temperature indication available to the plant operators. The RCS was subsequently pressurized to approximately 7 to 10 psig. The NRC AIT members concluded that the Diablo Canyon, Unit 2 plant was, at the time of the event, in a condition not previously analyzed by the NRC staff. The AIT findings from this event appear significant and generic to other pressurized water reactor facilities licensed by the NRC

  9. Air temperature determination inside residual heat removal pump room of Angra-1 nuclear power plant after a design basic accident

    International Nuclear Information System (INIS)

    Siniscalchi, Marcio Rezende

    2005-01-01

    This work develops heat transfer theoretical models for determination of air temperature inside the Residual Heat Removal Pump Room of Angra 1 Nuclear Power Plant after a Design Basis Accident without forced ventilation. Two models had been developed. The differential equations are solved by analytical methods. A software in FORTRAN language are developed for simulations of temperature inside rooms for different geometries and materials. (author)

  10. Extension and application of a scaling technique for duplication of in-flight aerodynamic heat flux in ground test facilities

    NARCIS (Netherlands)

    Veraar, R.G.

    2009-01-01

    To enable direct experimental duplication of the inflight heat flux distribution on supersonic and hypersonic vehicles, an aerodynamic heating scaling technique has been developed. The scaling technique is based on the analytical equations for convective heat transfer for laminar and turbulent

  11. [Steam and air co-injection in removing TCE in 2D-sand box].

    Science.gov (United States)

    Wang, Ning; Peng, Sheng; Chen, Jia-Jun

    2014-07-01

    Steam and air co-injection is a newly developed and promising soil remediation technique for non-aqueous phase liquids (NAPLs) in vadose zone. In this study, in order to investigate the mechanism of the remediation process, trichloroethylene (TCE) removal using steam and air co-injection was carried out in a 2-dimensional sandbox with different layered sand structures. The results showed that co-injection perfectly improved the "tailing" effect compared to soil vapor extraction (SVE), and the remediation process of steam and air co-injection could be divided into SVE stage, steam strengthening stage and heat penetration stage. Removal ratio of the experiment with scattered contaminant area was higher and removal speed was faster. The removal ratios from the two experiments were 93.5% and 88.2%, and the removal periods were 83.9 min and 90.6 min, respectively. Steam strengthened the heat penetration stage. The temperature transition region was wider in the scattered NAPLs distribution experiment, which reduced the accumulation of TCE. Slight downward movement of TCE was observed in the experiment with TCE initially distributed in a fine sand zone. And such downward movement of TCE reduced the TCE removal ratio.

  12. Optimization of armour geometry and bonding techniques for tungsten-armoured high heat flux components

    International Nuclear Information System (INIS)

    Giniyatulin, R.N.; Komarov, V.L.; Kuzmin, E.G.; Makhankov, A.N.; Mazul, I.V.; Yablokov, N.A.; Zhuk, A.N.

    2002-01-01

    Joining of tungsten with copper-based cooling structure and armour geometry optimization are the major aspects in development of the tungsten-armoured plasma facing components (PFC). Fabrication techniques and high heat flux (HHF) tests of tungsten-armoured components have to reflect different PFC designs and acceptable manufacturing cost. The authors present the recent results of tungsten-armoured mock-ups development based on manufacturing and HHF tests. Two aspects were investigated--selection of armour geometry and examination of tungsten-copper bonding techniques. Brazing and casting tungsten-copper bonding techniques were used in small mock-ups. The mock-ups with armour tiles (20x5x10, 10x10x10, 20x20x10, 27x27x10) mm 3 in dimensions were tested by cyclic heat fluxes in the range of (5-20) MW/m 2 , the number of thermal cycles varied from hundreds to several thousands for each mock-up. The results of the tests show the applicability of different geometry and different bonding technique to corresponding heat loading. A medium-scale mock-up 0.6-m in length was manufactured and tested. HHF tests of the medium-scale mock-up have demonstrated the applicability of the applied bonding techniques and armour geometry for full-scale PFC's manufacturing

  13. Radiation detector system having heat pipe based cooling

    Science.gov (United States)

    Iwanczyk, Jan S.; Saveliev, Valeri D.; Barkan, Shaul

    2006-10-31

    A radiation detector system having a heat pipe based cooling. The radiation detector system includes a radiation detector thermally coupled to a thermo electric cooler (TEC). The TEC cools down the radiation detector, whereby heat is generated by the TEC. A heat removal device dissipates the heat generated by the TEC to surrounding environment. A heat pipe has a first end thermally coupled to the TEC to receive the heat generated by the TEC, and a second end thermally coupled to the heat removal device. The heat pipe transfers the heat generated by the TEC from the first end to the second end to be removed by the heat removal device.

  14. Assessment of alternate ion exchange resins for improved antimony removal from the primary heat transport system

    Energy Technology Data Exchange (ETDEWEB)

    Burany, R.; Suryanarayan, S.; Husain, A. [Kinectrics, Inc., Toronto, ON (Canada)

    2015-07-01

    Radiation fields around the CANDU heat transport system are a major contributor to worker dose during inspection, maintenance and refurbishment activities. While Co-60 is typically the dominant contributor to radiation fields in CANDU reactors, Sb-124, an activation product of antimony, is also a significant contributor, accounting for 5-20% of the radiation fields. The goal of this research project was to investigate resins for improved removal of antimony under both oxidizing and reducing conditions.Several candidate resins were tested and short-listed through a sequence of iterative testing. The results of the laboratory testing have identified potential candidates for improved antimony removal. Further testing is required to ensure compatibility with existing station resin specifications. (author)

  15. Steady-state heat and particle removal with the actively cooled Phase III outboard pump limiter in Tore Supra

    International Nuclear Information System (INIS)

    Nygren, R.; Koski, J.; Lutz, T.; McGrath; Miller, J.; Watkins, J.; Guilhem, D.; Chappuis, P.; Cordier, J.; Loarer, T.

    1995-01-01

    Tore Supra's Phase III outboard pump limiter (OPL) is a modular actively-cooled mid-plane limiter, designed for heat and particle removal during long pulse operation. During its initial operation in 1993, the OPL successfully removed about 1 MW of power during ohmicly heated shots of up to 10 s duration and reached (steady state) thermal equilibrium. The particle pumping of the Phase III OPL was found to be about 50% greater than the Phase II OPL which had a radial distance between the last closed flux surface and the entrance of the pumping throat of 3.5 cm compared with only 2.5 cm for the Phase III OPL. This paper gives examples of power distribution over the limiter from IR measurements of surface temperature and from extensively calorimetry (34 thermocouples and 10 flow meters) and compares the distributions with values predicted by a 3D model (HF3D) with a detailed magnetic configuration (e.g., includes field ripple). ((orig.))

  16. Treatment techniques for removing natural radionuclides from drinking water. Final report of the TENAWA project

    International Nuclear Information System (INIS)

    Annanmaeki, M.; Turtiainen, T.

    2000-01-01

    TENAWA project (Treatment Techniques for Removing Natural Radionuclides from Drinking Water) was carried out on a cost-shared basic with the European Commission (CEC) under the supervision of Directorate-General XII, Radiation Protection Unit. TENAWA project was started because in several European countries ground water supplies may contain high amounts of natural radionuclides. During the project both laboratory and field research was performed in order to test the applicability of different equipment and techniques for removing natural radionuclides from drinking water. The measurable objectives of the project were: to give recommendations on the most suitable methods for removing radon ( 222 Rn), uranium ( 238,234 U), radium ( 226 , 228 Ra), lead ( 210 Pb) and polonium ( 210 Po) from drinking water of different qualities (i.e. soft, hard, iron-, manganese- and humus-rich, acidic) to test commercially available equipment for its ability to remove radionuclides; to find new materials, absorbents and membranes effective in the removal of radionuclides and to issue guidelines for the treatment and disposal of radioactive wastes produced in water treatment. Radon could be removed efficiently (>95%) from domestic water supplies by both aeration and granular activated carbon (GAC) filtration. Defects in technical reliability or radon removal efficiency were observed in some aerators. The significant drawback of GAC filtration was the elevated gamma dose rates (up to 120 μSv/h) near the filter and the radioactivity of spent GAC. Aeration was found to be a suitable method for removing radon at waterworks, too. The removal efficiencies at waterworks where the aeration process was designed to remove radon or carbon dioxide were 67-99%. If the aeration process was properly designed, removal efficiencies higher than 95% could be attained. Uranium could best be removed (>95%) with strong basic anion exchange resins and radium by applying strong acidic cation exchange resins

  17. Treatment techniques for removing natural radionuclides from drinking water. Final report of the TENAWA project

    Energy Technology Data Exchange (ETDEWEB)

    Annanmaeki, M.; Turtiainen, T. [eds.

    2000-01-01

    TENAWA project (Treatment Techniques for Removing Natural Radionuclides from Drinking Water) was carried out on a cost-shared basic with the European Commission (CEC) under the supervision of Directorate-General XII, Radiation Protection Unit. TENAWA project was started because in several European countries ground water supplies may contain high amounts of natural radionuclides. During the project both laboratory and field research was performed in order to test the applicability of different equipment and techniques for removing natural radionuclides from drinking water. The measurable objectives of the project were: to give recommendations on the most suitable methods for removing radon ({sup 222}Rn), uranium ({sup 238,234}U), radium ({sup 226}, {sup 228}Ra), lead ({sup 210}Pb) and polonium ({sup 210}Po) from drinking water of different qualities (i.e. soft, hard, iron-, manganese- and humus-rich, acidic) to test commercially available equipment for its ability to remove radionuclides; to find new materials, absorbents and membranes effective in the removal of radionuclides and to issue guidelines for the treatment and disposal of radioactive wastes produced in water treatment. Radon could be removed efficiently (>95%) from domestic water supplies by both aeration and granular activated carbon (GAC) filtration. Defects in technical reliability or radon removal efficiency were observed in some aerators. The significant drawback of GAC filtration was the elevated gamma dose rates (up to 120 {mu}Sv/h) near the filter and the radioactivity of spent GAC. Aeration was found to be a suitable method for removing radon at waterworks, too. The removal efficiencies at waterworks where the aeration process was designed to remove radon or carbon dioxide were 67-99%. If the aeration process was properly designed, removal efficiencies higher than 95% could be attained. Uranium could best be removed (>95%) with strong basic anion exchange resins and radium by applying strong

  18. Modelling of decay heat removal using large water pools

    International Nuclear Information System (INIS)

    Munther, R.; Raussi, P.; Kalli, H.

    1992-01-01

    The main task for investigating of passive safety systems typical for ALWRs (Advanced Light Water Reactors) has been reviewing decay heat removal systems. The reference system for calculations has been represented in Hitachi's SBWR-concept. The calculations for energy transfer to the suppression pool were made using two different fluid mechanics codes, namely FIDAP and PHOENICS. FIDAP is based on finite element methodology and PHOENICS uses finite differences. The reason choosing these codes has been to compare their modelling and calculating abilities. The thermal stratification behaviour and the natural circulation was modelled with several turbulent flow models. Also, energy transport to the suppression pool was calculated for laminar flow conditions. These calculations required a large amount of computer resources and so the CRAY-supercomputer of the state computing centre was used. The results of the calculations indicated that the capabilities of these codes for modelling the turbulent flow regime are limited. Output from these codes should be considered carefully, and whenever possible, experimentally determined parameters should be used as input to enhance the code reliability. (orig.). (31 refs., 21 figs., 3 tabs.)

  19. Simple Technique for Removing Broken Pedicular Screws

    Directory of Open Access Journals (Sweden)

    A Agrawal

    2014-03-01

    Full Text Available The procedure for removing a broken pedicle screw should ideally be technically easy and minimally invasive, as any damage to the pedicle, during removal of the broken screw, may weaken the pedicle, thus compromising on the success of re-instrumentation. We describe the case of a 32-year old man who had undergone surgery for traumatic third lumbar vertebral body fracture three years prior to current admission and had developed the complication of pedicle screw breakage within the vertebral body. The patient underwent re-exploration and removal of the distal screws. Through a paravertebral incision and muscle separation, the screws and rods were exposed and the implants were removed.

  20. Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe

    International Nuclear Information System (INIS)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol

    2015-01-01

    Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of

  1. Analytical studies on the impact of using repeated-rib roughness in LMR [Liquid Metal Reactor] decay heat removal systems

    International Nuclear Information System (INIS)

    Obot, N.T.; Tessier, J.H.; Pedersen, D.R.

    1988-01-01

    A numerical study was carried out to determine the effects of roughness on the thermal performance of Liquid Metal Reactor (LMR) decay heat removal systems for a range of possible design configurations and operating conditions. The ranges covered for relative rib height (e/D/sub h/), relative pitch (p/e) and flow attack angle were 0.026--0.103, 5--20 and 0--90 degrees, successively. The heat flux was varied between 1.1 and 21.5 kW/m 2 (0.1 and 2.0 kW/ft 2 ). Calculations were made for three cases: smooth duct with no ribs, ribs on both the guard vessel and collector wall, and ribs on the collector wall only. The results indicate that significant benefits, amounting to nearly two-fold reductions in guard vessel and collector wall temperatures, can be realized by placing repeated ribs on both the guard vessel and the collector wall. The magnitudes of the reduction in the reactor vessel temperature are considerably smaller. In general, the level of improvement, be it with respect to temperature or heat flux, is only mildly affected by changes in rib height or pitch but exhibits greater sensitivity to the assumed value for the system form loss. When the ribs are placed only on the collector wall, the heat removal capability is substantially reduced

  2. Experiences with on line fault detection system for protection system logic and decay heat removal system logic in Dhruva

    International Nuclear Information System (INIS)

    Ramkumar, N.; Dutta, P.K.; Darbhe, M.D.; Bharadwaj, G.

    2001-01-01

    Dhruva is a 100 MW (Thermal) natural uranium fuelled, vertical core, tank type multi purpose research reactor with heavy water acting as moderator, coolant and reflector. Helium is used as cover gas for heavy water system. Reactor Protection System and Decay Heat Removal System (DHRS) have triplicated instrumented channels. The logic for these systems are hybrid in nature with a mixture of relay logic and solid state logic. Fine Impulse Technique(FIT) is employed for On-line fault detection in the solid state logics of these systems. The FIT systems were designed in the early eighties. Operating experiences over the past 15 years has revealed certain deficiencies. In view of this, a microcomputer based state of the art FIT systems for logics of Reactor Protection System and DHRS are being implemented with improved functionalities built into them. This paper describes the operating experience of old FIT systems and improved features of the proposed new FIT systems. (author)

  3. Pattern transformation of heat-shrinkable polymer by three-dimensional (3D) printing technique.

    Science.gov (United States)

    Zhang, Quan; Yan, Dong; Zhang, Kai; Hu, Gengkai

    2015-03-11

    A significant challenge in conventional heat-shrinkable polymers is to produce controllable microstructures. Here we report that the polymer material fabricated by three-dimensional (3D) printing technique has a heat-shrinkable property, whose initial microstructure can undergo a spontaneous pattern transformation under heating. The underlying mechanism is revealed by evaluating internal strain of the printed polymer from its fabricating process. It is shown that a uniform internal strain is stored in the polymer during the printing process and can be released when heated above its glass transition temperature. Furthermore, the internal strain can be used to trigger the pattern transformation of the heat-shrinkable polymer in a controllable way. Our work provides insightful ideas to understand a novel mechanism on the heat-shrinkable effect of printed material, but also to present a simple approach to fabricate heat-shrinkable polymer with a controllable thermo-structural response.

  4. Piezosurgery for the lingual split technique in mandibular third molar removal: a suggestion.

    Science.gov (United States)

    Pippi, Roberto; Alvaro, Roberto

    2013-03-01

    The lingual split technique is a surgical procedure for extraction of impacted mandibular third molar throughout a lingual approach. The main disadvantage of this technique is the high rate of temporary lingual nerve injury mainly because of the trauma induced by the lingual flap retraction. The purpose of this paper is to suggest the use of piezosurgery in performing the lingual cortical plate osteotomy of the third molar alveolar process. Surgical procedure was performed under general anesthesia, and it lasted approximately 60 minutes. After the buccal and lingual full-thickness flaps were incised and elevated, a piezosurgical device was used for osteotomy. A well-defined bony window was then removed, and it allowed the entire tooth was extracted in a lingual direction. The patient did not show any neurological postoperative complication. Lingual and inferior alveolar nerve functionality was normal before as well as after surgery. The use of piezoelectric surgery seems to be a good option in removing lower third molars when a lingual access is clearly indicated. The only disadvantage of this technique can be represented by an operating time lengthening possibly because of a lower power cut of the piezoelectric device, to the high mineralization of the mandibular cortical bone and to the use of inserts with a low degree of sharpening.

  5. Conceptual design of cesium removal device for ITER NBI maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Shibanuma, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Cesium is required in order to generate a stable negative ion of hydrogen in an ion source of the neutral beam injector (NBI), which is one of the plasma-heating devices for International Thermonuclear Experimental Reactor (ITER). After long time operation of the NBI, the cesium deposits to the insulators supporting the electrode. Due to the deterioration of the insulation resistance, the continuous operation of the NBI will be difficult. In addition, the NBI device is activated by neutrons from D-T plasma, so that periodic removal and cleaning of the cesium on the insulators by remove handling is required. A study of the cesium removal scenario and the device is therefore required considering remote handling. In this report, a cesium removal procedure and conceptual design of the cesium removal device using laser ablation technique are studied, and the feasibility of the laser ablation method is shown. (author)

  6. Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Bang, In Cheol

    2017-01-01

    Highlights: • Thermal performances and operation limits of hybrid heat pipe were experimentally studied. • Models for predicting the operation limit of the hybrid heat pipe was developed. • Non-condensable gas affected heat transfer characteristics of the hybrid heat pipe. - Abstract: In this paper, a hybrid heat pipe is proposed for use in advanced nuclear power plants as a passive heat transfer device. The hybrid heat pipe combines the functions of a heat pipe and a control rod to simultaneously remove the decay heat generated from the core and shutdown the reactor under accident conditions. Thus, the hybrid heat pipe contains a neutron absorber in the evaporator section, which corresponds to the core of the reactor pressure vessel. The presence of the neutron absorber material leads to differences in the heated diameter and hydraulic diameter of the heat pipe. The cross-sectional areas of the vapor paths through the evaporator, adiabatic, and condenser sections are also different. The hybrid heat pipe must operate in a high-temperature, high-pressure environment to remove the decay heat. In other words, the operating pressure must be higher than those of the commercially available thermosyphons. Hence, the thermal performances, including operation limit of the hybrid heat pipe, were experimentally studied in the operating pressure range of 0.2–20 bar. The operating pressure of the hybrid heat pipe was controlled by charging the non-condensable gas which is unused method to achieve the high saturation pressure in conventional thermosyphons. The effect of operating pressure on evaporation heat transfer was negligible, while condensation heat transfer was affected by the amount of non-condensable gas in the test section. The operation limit of the hybrid heat pipe increased with the operating pressure. Maximum heat removal capacity of the hybrid heat pipe was up to 6 kW which is meaningful value as a passive decay heat removal device in the nuclear power

  7. Ghost fringe removal techniques using Lissajous data presentation.

    Science.gov (United States)

    Erskine, David J; Eggert, J H; Celliers, P M; Hicks, D G

    2016-03-01

    A VISAR (Velocity Interferometer System for Any Reflector) is a Doppler velocity interferometer which is an important optical diagnostic in shockwave experiments at the national laboratories, used to measure equation of state (EOS) of materials under extreme conditions. Unwanted reflection of laser light from target windows can produce an additional component to the VISAR fringe record that can distort and obscure the true velocity signal. Accurately removing this so-called ghost artifact component is essential for achieving high accuracy EOS measurements, especially when the true light signal is only weakly reflected from the shock front. Independent of the choice of algorithm for processing the raw data into a complex fringe signal, we have found it beneficial to plot this signal as a Lissajous and seek the proper center of this path, even under time varying intensity which can shift the perceived center. The ghost contribution is then solved by a simple translation in the complex plane that recenters the Lissajous path. For continuous velocity histories, we find that plotting the fringe magnitude vs nonfringing intensity and optimizing linearity is an invaluable tool for determining accurate ghost offsets. For discontinuous velocity histories, we have developed graphically inspired methods which relate the results of two VISARs having different velocity per fringe proportionalities or assumptions of constant fringe magnitude to find the ghost offset. The technique can also remove window reflection artifacts in generic interferometers, such as in the metrology of surfaces.

  8. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  9. Evaluation of effectiveness of cement removal from implant-retained crowns using a proposed circular crisscross flossing technique.

    Science.gov (United States)

    Ferreira, Cimara Fortes; Shafter, Mohamed Amer; Jain, Vinay; Wicks, Russel Anthony; Linder, Erno; Ledo, Carlos Alberto da Silva

    2018-02-13

    Extruded cement during dental implant crown cementation may cause peri-implant diseases if not removed adequately. Evaluate the efficiency of removal of cement after cementation of implant crowns using an experimental "circular crisscross flossing technique (CCCFT) flossing technique, compared to the conventional "C" shape flossing technique (CSFT). Twenty-four patients rendered 29 experimental and 29 control crowns. Prefabricated abutments were secured to the implant with the margins at least 1 mm subgingivally. The abutments were scanned using CADCAM technology and Emax crowns were fabricated in duplicates. Each crown was cemented separately and excess cement was removed using the CSFT and the CCFT techniques. After completion of cementation was completed, the screw access holes were accessed and the crown was unscrewed along with the abutment. The samples were disinfected using 70% ethanol for 10 minutes. Crowns were divided into 4 parts using a marker in order to facilitate measurement data collection. Vertical and horizontal measurements were made for extruded cement for each control and experimental groups by means of a digital microscope. One-hundred and seventeen measurements were made for each group. Mann-Whitney test was applied to verify statistical significance between the groups. The CCFT showed a highly statistically significant result (104.8 ± 13.66, pcrowns cementation when compared with the CSFT.

  10. Nuclear reactor equipped with a flooding tank and a residual heat removal and emergency cooling system

    International Nuclear Information System (INIS)

    Schabert, H.P.; Winkler, F.

    1975-01-01

    A description is given of a nuclear reactor such as a pressurized-water reactor or the like which is equipped with a flooding tank and a residual heat removal and emergency cooling system. The flooding tank is arranged within the containment shell at an elevation above the upper edge of the reactor core and contains a liquid for flooding the reactor core in the event of a loss of coolant

  11. Heat-Assisted Machining for Material Removal Improvement

    Science.gov (United States)

    Mohd Hadzley, A. B.; Hafiz, S. Muhammad; Azahar, W.; Izamshah, R.; Mohd Shahir, K.; Abu, A.

    2015-09-01

    Heat assisted machining (HAM) is a process where an intense heat source is used to locally soften the workpiece material before machined by high speed cutting tool. In this paper, an HAM machine is developed by modification of small CNC machine with the addition of special jig to hold the heat sources in front of the machine spindle. Preliminary experiment to evaluate the capability of HAM machine to produce groove formation for slotting process was conducted. A block AISI D2 tool steel with100mm (width) × 100mm (length) × 20mm (height) size has been cut by plasma heating with different setting of arc current, feed rate and air pressure. Their effect has been analyzed based on distance of cut (DOC).Experimental results demonstrated the most significant factor that contributed to the DOC is arc current, followed by the feed rate and air pressure. HAM improves the slotting process of AISI D2 by increasing distance of cut due to initial cutting groove that formed during thermal melting and pressurized air from the heat source.

  12. Process of heat exchangers manufactured by Scientific and Engineering Services (S.E.S.)

    International Nuclear Information System (INIS)

    Khurshid, S.

    1995-01-01

    The objective of this report is to have a comprehensive overview of heat exchanger fouling problems as they occur in the industrial manufacturing sector. Specifically the types of fouling, conditions that influence fouling, the mitigation and accommodation techniques currently being used and mitigation technology trends are discussed. Finally a table top apparatus is designed and fabricated to develop an on line mechanical technique of rubber ball system for the removal of dirt and scales of that exchangers. Experimental results of this apparatus show that the cleaning efficiency can be raised if this mechanical technique is used together with current chemical fouling removal methods. (author)

  13. Studies for the use of water soluble chelating polymer in ultra-filtration technique for the removal of uranium from aqueous solutions

    International Nuclear Information System (INIS)

    Misra, S.K.; Mahatele, A.K.; Tripathi, S.C.; Vijayan, K.; Munshi, S.K.

    2005-01-01

    Studies were carried out for the removal of uranium from aqueous medium using water soluble chelating polymer by ultra-filtration technique. The water soluble polymers are the option for the surfactants used in the micellar enhanced ultra-filtration technique. More than 95% separation of uranium carried out under different experimental conditions, suggest that the technique can be effectively employed for the removal uranium from the aqueous effluent streams. (author)

  14. Microbiological analysis after complete or partial removal of carious dentin using two different techniques in primary teeth: A randomized clinical trial

    Science.gov (United States)

    Singhal, Deepak Kumar; Acharya, Shashidhar; Thakur, Arun Singh

    2016-01-01

    Background: The management of deep carious lesions can be done by various techniques but residual caries dilemma still persists and bacterial reduction in cavities treated by either partial or complete caries removal techniques is debatable. So the objective of the present randomized clinical trial was to compare microbial counts in cavities submitted to complete caries removal and partial caries removal using either hand instruments or burs before and after 3 weeks of restoration. Materials and Methods: Primary molars with acute carious lesions in inner half of dentine and vital pulp were randomly divided into three groups of 14 each: Group A: Partial caries removal using hand instruments atraumatic restorative treatment (ART) only; Group B: Partial caries removal using bur; Group C: Complete caries removal using bur and caries detector dye. Dentine sample obtained after caries removal and 3 weeks after restoration, were subjected to microbial culture and counting (colony-forming units [CFU]/mg of dentine) for total viable bacterial count, Streptococcus spp., mutans streptococci, Lactobacillus spp. Results: Three techniques of caries removal showed significant (P < 0.05) reduction in all microorganisms studied after 3 weeks of evaluation, but there was no statistically significant difference in percentage reduction of microbial count among three groups. Conclusion: Results suggest the use of partial caries removal in a single session as compared to complete caries removal as a part of treatment of deep lesions in deciduous teeth in order to reduce the risk of pulp exposure. Partial caries removal using ART can be preferred for community settings as public health procedure for caries management. PMID:26962313

  15. Analyses of Effects of Cutting Parameters on Cutting Edge Temperature Using Inverse Heat Conduction Technique

    Directory of Open Access Journals (Sweden)

    Marcelo Ribeiro dos Santos

    2014-01-01

    Full Text Available During machining energy is transformed into heat due to plastic deformation of the workpiece surface and friction between tool and workpiece. High temperatures are generated in the region of the cutting edge, which have a very important influence on wear rate of the cutting tool and on tool life. This work proposes the estimation of heat flux at the chip-tool interface using inverse techniques. Factors which influence the temperature distribution at the AISI M32C high speed steel tool rake face during machining of a ABNT 12L14 steel workpiece were also investigated. The temperature distribution was predicted using finite volume elements. A transient 3D numerical code using irregular and nonstaggered mesh was developed to solve the nonlinear heat diffusion equation. To validate the software, experimental tests were made. The inverse problem was solved using the function specification method. Heat fluxes at the tool-workpiece interface were estimated using inverse problems techniques and experimental temperatures. Tests were performed to study the effect of cutting parameters on cutting edge temperature. The results were compared with those of the tool-work thermocouple technique and a fair agreement was obtained.

  16. Deep underground reactor (passive heat removal of LWR with hard neutron energy spectrum)

    Energy Technology Data Exchange (ETDEWEB)

    Hiroshi, Takahashi [Brookhaven National Lab., Upton, NY (United States)

    2001-07-01

    To run a high conversion reactor with Pu-Th fueled tight fueled assembly which has a long burn-up of a fuel, the reactor should be sited deep underground. By putting the reactor deep underground heat can be removed passively not only during a steady-state run and also in an emergency case of loss of coolant and loss of on-site power; hence the safety of the reactor can be much improved. Also, the evacuation area around the reactor can be minimized, and the reactor placed near the consumer area. This approach reduces the cost of generating electricity by eliminating the container building and shortening transmission lines. (author)

  17. Deep underground reactor (passive heat removal of LWR with hard neutron energy spectrum)

    International Nuclear Information System (INIS)

    Hiroshi, Takahashi

    2001-01-01

    To run a high conversion reactor with Pu-Th fueled tight fueled assembly which has a long burn-up of a fuel, the reactor should be sited deep underground. By putting the reactor deep underground heat can be removed passively not only during a steady-state run and also in an emergency case of loss of coolant and loss of on-site power; hence the safety of the reactor can be much improved. Also, the evacuation area around the reactor can be minimized, and the reactor placed near the consumer area. This approach reduces the cost of generating electricity by eliminating the container building and shortening transmission lines. (author)

  18. Development of technique to apply induction heating stress improvement to recirculation inlet nozzle

    International Nuclear Information System (INIS)

    Chiba, Kunihiko; Nihei, Kenichi; Ootaka, Minoru

    2009-01-01

    Stress corrosion cracking (SCC) have been found in the primary loop recirculation (PLR) systems of boiling water reactors (BWR). Residual stress in welding heat-affected zone is one of the factors of SCC, and the residual stress improvement is one of the most effective methods to prevent SCC. Induction heating stress improvement (IHSI) is one of the techniques to improve reduce residual stress. However, it is difficult to apply IHSI to the place such as the recirculation inlet nozzle where the flow stagnates. In this present study, the technique to apply IHSI to the recirculation inlet nozzle was developed using water jet which blowed into the crevice between the nozzle safe end and the thermal sleeve. (author)

  19. Atomic Oxygen Treatment Technique for Removal of Smoke Damage from Paintings

    Science.gov (United States)

    Rutledge, S. K.; Banks, B. A.

    1997-01-01

    Soot deposits that can accumulate on surfaces of a painting during a fire can be difficult to clean from some types of paintings without damaging the underlying paint layers. A non-contact technique has been developed which can remove the soot by allowing a gas containing atomic oxygen to flow over the surface and chemically react with the soot to form carbon monoxide and carbon dioxide. The reaction is limited to the surface, so the underlying paint is not touched. The process can be controlled so that the cleaning can be stopped once the paint surface is reached. This paper describes the smoke exposure and cleaning of untreated canvas, acrylic gesso, and sections of an oil painting using this technique. The samples were characterized by optical microscopy and reflectance spectroscopy.

  20. Dynamic simulation of the air-cooled decay heat removal system of the German KNK-II experimental breeder reactor

    International Nuclear Information System (INIS)

    Schubert, B.K.

    1984-07-01

    A Dump Heat Exchanger and associated feedback control system models for decay heat removal in the German KNK-II experimental fast breeder reactor are presented. The purpose of the controller is to minimize temperature variations in the circuits and, hence, to prevent thermal shocks in the structures. The basic models for the DHX include the sodium-air thermodynamics and hydraulics, as well as a control system. Valve control models for the primary and intermediate sodium flow regulation during post shutdown conditions are also presented. These models have been interfaced with the SSC-L code. Typical results of sample transients are discussed

  1. Heat Exchanger Tube Inspection of Nuclear Power Plants using IRIS Technique

    International Nuclear Information System (INIS)

    Yoon, Byung Sik; Yang, Seung Han; Song, Seok Yoon; Kim, Yong Sik; Lee, Hee Jong

    2005-01-01

    Inspection of heat exchange tubing include steam generator of nuclear power plant mostly performed with eddy current method. Recently, various inspection technique is available such as remote field eddy current, flux leakage and ultrasonic methods. Each of these techniques has its merits and limitations. Electromagnetic techniques are very useful to locate areas of concern but sizing is hard because of the difficult interpretation of an electric signature. On the other hand, ultrasonic methods are very accurate in measuring wall loss damage, and are reliable for detecting cracks. Additionally ultrasound methods is not affected by support plates or tube sheets and variation of electrical conductivity or permeability. Ultrasound data is also easier to analyze since the data displayed is generally the remaining wall thickness. It should be emphasized that ultrasound is an important tool for sizing defects in tubing. In addition, it can be used in situations where eddy current or remote field eddy current is not reliable, or as a flaw assessment tool to supplement the electromagnetic data. The need to develop specialized ultrasonic tools for tubing inspection was necessary considering the limitations of electromagnetic techniques to some common inspection problems. These problems the sizing of wall loss in carbon steel tubes near the tube sheet or support plate, sizing internal erosion damage, and crack detection. This paper will present an IRIS(Internal Rotating Inspection System) ultrasonic tube inspection technique for heat exchanger tubing in nuclear power plant and verify inspection reliability for artificial flaw embedded to condenser tube

  2. Studies on the characteristics of the separated type heat pipe system with non-condensible gas for the use of the passive decay heat removal in reactor systems

    International Nuclear Information System (INIS)

    Hayashi, Takao; Iigaki, Kazuhiko; Ohashi, Kazutaka; Hayakawa, Hitoshi; Yamada, Masao.

    1995-01-01

    This study is the fundamental research by experiments to aim at the development of the complete passive decay heat removal system on the modular reactor systems by the form of the separated type of heat pipe system utilizing the features of both the big latent heat for vaporization from water to steam and easy transportation characteristics. Special intention in our study on the fundamental experiments is to look for the effects in such a separated type of heat pipe system to introduce non-condensible gas such as nitrogen gas together with the working fluid of water. Many interesting findings have been obtained so far on the experiments for the variable conductance heat pipe characteristics from viewpoint of the actual application on the aim said above. This study has been carried out by the joint study between Tokai University and Fuji Electric Co., Ltd. and this paper is made up from the several papers presented so far at both the national and international symposiums under the name of joint study of the both bodies. (author)

  3. Removing lead from metallic mixture of waste printed circuit boards by vacuum distillation: factorial design and removal mechanism.

    Science.gov (United States)

    Li, Xingang; Gao, Yujie; Ding, Hui

    2013-10-01

    The lead removal from the metallic mixture of waste printed circuit boards by vacuum distillation was optimized using experimental design, and a mathematical model was established to elucidate the removal mechanism. The variables studied in lead evaporation consisted of the chamber pressure, heating temperature, heating time, particle size and initial mass. The low-level chamber pressure was fixed at 0.1 Pa as the operation pressure. The application of two-level factorial design generated a first-order polynomial that agreed well with the data for evaporation efficiency of lead. The heating temperature and heating time exhibited significant effects on the efficiency, which was validated by means of the copper-lead mixture experiments. The optimized operating conditions within the region studied were the chamber pressure of 0.1 Pa, heating temperature of 1023 K and heating time of 120 min. After the conditions were employed to remove lead from the metallic mixture of waste printed circuit boards, the efficiency was 99.97%. The mechanism of the effects was elucidated by mathematical modeling that deals with evaporation, mass transfer and condensation, and can be applied to a wider range of metal removal by vacuum distillation. Copyright © 2013 Elsevier Ltd. All rights reserved.

  4. Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

    1997-07-01

    Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

  5. Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Hu, Rui; Lisowski, Darius; Kraus, Adam

    2016-04-17

    The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at the NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.

  6. Heat transfer monitoring by means of the hot wire technique and finite element analysis software.

    Science.gov (United States)

    Hernández Wong, J; Suarez, V; Guarachi, J; Calderón, A; Rojas-Trigos, J B; Juárez, A G; Marín, E

    2014-01-01

    It is reported the study of the radial heat transfer in a homogeneous and isotropic substance with a heat linear source in its axial axis. For this purpose, the hot wire characterization technique has been used, in order to obtain the temperature distribution as a function of radial distance from the axial axis and time exposure. Also, the solution of the transient heat transport equation for this problem was obtained under appropriate boundary conditions, by means of finite element technique. A comparison between experimental, conventional theoretical model and numerical simulated results is done to demonstrate the utility of the finite element analysis simulation methodology in the investigation of the thermal response of substances. Copyright © 2013 Elsevier Ltd. All rights reserved.

  7. Energy balance of droplets impinging onto a wall heated above the Leidenfrost temperature

    International Nuclear Information System (INIS)

    Dunand, P.; Castanet, G.; Gradeck, M.; Maillet, D.; Lemoine, F.

    2013-01-01

    Highlights: • Measurement techniques are combined to characterize the heat lost due to liquid vaporization. • The wall heat flux is estimated by infrared thermography associated with inverse heat conduction. • The liquid heating is characterized by the two-color Laser-Induced Fluorescence thermometry. • Results reveal how the heat fluxes vary with the droplet sizes and the Weber number. -- Abstract: This work is an experimental study aiming at characterizing the heat transfers induced by the impingement of water droplets (diameter 80–180 μm) on a thin nickel plate heated by electromagnetic induction. The temperature of the rear face of the nickel sample is measured by means of an infrared camera and the heat removed from the wall due to the presence of the droplets is estimated using a semi-analytical inverse heat conduction model. In parallel, the temperature of the droplets is measured using the two-color Laser-Induced Fluorescence thermometry (2cLIF) which has been extended to imagery for the purpose of these experiments. The measurements of the variation in the droplet temperature occurring during an impact allow determining the sensible heat removed by the liquid. Measurements are performed at wall conditions well above the Leidenfrost temperature. Different values of the Weber numbers corresponding to the bouncing and splashing regimes are tested. Comparisons between the heat flux removed from the wall and the sensible heat gained by the liquid allows estimating the heat flux related to liquid evaporation. Results reveal that the respective level of the droplet sensible heat and the heat lost due to liquid vaporization can vary significantly with the droplet sizes and the Weber number

  8. A new dewatering technique for stingless bees honey

    Directory of Open Access Journals (Sweden)

    Ramli Ahmad Syazwan

    2017-01-01

    Full Text Available One of the problems faced in stingless bee honey storage is spoilage by the fermentation process occurs in honey due to its high water content. There are a few techniques available currently, but they are time consuming and there is excessive heat involved in the process. The temperature of the process must be kept low because excessive heat can deteriorate nutrition value and biochemical content in honey. Hence, a new method of honey dewatering was developed using a Low Temperature Vacuum Drying (LTVD with induced nucleation technique.The objective of this research is to investigate the performance of a LTVD with induced nucleation to reduce the water content in honey. First, the honey was placed in a pressure vessel, and then air was removed. Then, the honey was slightly heated at 30°C and the water content before and after the experiment was measured by a refractometer. The steps were repeated until the water content reached below 20%. It was found that the LTVD method improved the water removal rate significantly with an average of 0.15% of water content per minute. That is 3 times much faster than the conventional method of low temperature heating by Tabouret. Higher temperature during dewatering process improved the dewatering rate significantly. It can be concluded that LTVD is a promising option in tackling the high water content in stingless bee honey issue.

  9. Containment heat removal system

    International Nuclear Information System (INIS)

    Wade, G.E.; Barbanti, G.; Gou, P.F.; Rao, A.S.; Hsu, L.C.

    1992-01-01

    This patent describes a nuclear system of a type including a containment having a nuclear reactor therein, the nuclear reactor including a pressure vessel and a core in the pressure vessel, the system. It comprises a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant into the nuclear reactor pressure vessel against a predetermined pressure within the nuclear reactor pressure vessel; means for reducing a pressure of steam in the nuclear reactor pressure vessel to a value less than the predetermined pressure in the event of a nuclear accident, the means including a depressurization valve connected to the pressure vessel, the means further including steam heat dissipating means such dissipating means including a suppression pool; a supply of water in the suppression pool, there being a headspace in the suppression pool above the water supply; a substantial amount of air in the head space; means for feeding pressurized steam from the nuclear reactor pressure vessel to a location under a surface of the supply of water, the supply of water being effective to absorb heat sufficient to reduce steam pressure below the predetermined pressure; and a check valve for communicating the headspace with the containment, the check valve being oriented to vent air in the headspace to the containment when a pressure in the headspace exceeds a pressure in the containment by a predetermined pressure differential

  10. Effect of Simplifying Drilling Technique on Heat Generation During Osteotomy Preparation for Dental Implant.

    Science.gov (United States)

    El-Kholey, Khalid E; Ramasamy, Saravanan; Kumar R, Sheetal; Elkomy, Aamna

    2017-12-01

    To test the hypothesis that there would be no difference in heat production by reducing the number of drills during the implant site preparation relative to conventional drilling sequence. A total of 120 implant site preparations with 3 different diameters (3.6, 4.3, and 4.6 mm) were performed on bovine ribs. Within the same diameter group, half of the preparations were performed by a simplified drilling procedure (pilot drill + final diameter drill) and other half using the conventional drilling protocol (pilot drill followed by graduated series of drills to widen the site). Heat production by different drilling techniques was evaluated by measuring the bone temperature using k-type thermocouple and a sensitive thermometer before and after each drill. Mean for maximum temperature increase during site preparation of the 3.6, 4.3, and 4.6-mm implants was 2.45, 2.60, and 2.95° when the site was prepared by the simplified procedure, whereas it was 2.85, 3.10, and 3.60° for the sites prepared by the conventional technique, respectively. No significant difference in temperature increase was found when implants of the 3 different diameters were prepared either by the conventional or simplified drilling procedure. The simplified drilling technique produced similar amount of heat comparable to the conventional technique that proved the initial hypothesis.

  11. Heat transfer enhancement

    International Nuclear Information System (INIS)

    Hasatani, Masanobu; Itaya, Yoshinori

    1985-01-01

    In order to develop energy-saving techniques and new energy techniques, and also most advanced techniques by making industrial equipment with high performance, heat transfer performance frequently becomes an important problem. In addition, the improvement of conventional heat transfer techniques and the device of new heat transfer techniques are often required. It is most proper that chemical engineers engage in the research and development for enhancing heat transfer. The research and development for enhancing heat transfer are important to heighten heat exchange efficiency or to cool equipment for preventing overheat in high temperature heat transfer system. In this paper, the techniques of enhancing radiative heat transfer and the improvement of radiative heat transfer characteristics are reported. Radiative heat transfer is proportional to fourth power of absolute temperature, and it does not require any heat transfer medium, but efficient heat-radiation converters are necessary. As the techniques of enhancing radiative heat transfer, the increase of emission and absorption areas, the installation of emissive structures and the improvement of radiative characteristics are discussed. (Kako, I.)

  12. Green polyurethane synthesis by emulsion technique: a magnetic composite for oil spill removal

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Raphael Maria Dias da; Hungerbühler, Gabriela; Saraiva, Thiago; De Jong, Gabriel; Moraes, Rafael Silva; Ferreira, Luciana Spinelli; Souza Junior, Fernando Gomes de, E-mail: fgsj@ufrj.br [Universidade Federal do Rio de Janeiro (LaBioS/UFRJ), Rio de Janeiro, RJ (Brazil). Lab. de Biopolímeros e Sensores; Furtado, Evandro Goncalves [Alfa Rio Química Ltda., Duque de Caxias, RJ (Brazil); Silva, Fabrício Machado [Universidade de Brasília (UnB), Brasília, DF (Brazil); Oliveira, Geiza Esperandio de [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    After the consolidation of the Brazilian biodiesel industry, issues related to the final destination of the glycerin, the by-product from the biodiesel industrial process, drawing the attention of several researchers. There are several uses to this byproduct. Among them, the obtaining of polymers, such as polyurethane (PU), are very encouraged since the glycerin ca be used, as well as the castor oil, in the replacement of petrochemical polyols. The aim of this work was to propose a new route for the obtainment of a petroleum sorbent based on polyurethane resin from glycerin and castor oil, through the emulsion technique. In addition, maghemite (γ-Fe{sub 2}O{sub 3}) was mixed to the polymer matrix, producing a magnetic composite, able to make easier the oil cleanup process. The products synthesized were characterized by Fourier transform infrared spectroscopy, X-ray diffraction, simultaneous Thermogravimetry (TGA) and Differential scanning calorimetry (DSC), Optical microscopy, Scanning electron microscopy (SEM). In addition, magnetic force and oil removal capability tests were also performed. The magnetic material was used to remove oil from water, exhibited a good oil removal capability. In a typical test, 1g of the composite containing 5wt% of maghemite was able to remove 10g of oil from water. (author)

  13. Green polyurethane synthesis by emulsion technique: a magnetic composite for oil spill removal

    International Nuclear Information System (INIS)

    Costa, Raphael Maria Dias da; Hungerbühler, Gabriela; Saraiva, Thiago; De Jong, Gabriel; Moraes, Rafael Silva; Ferreira, Luciana Spinelli; Souza Junior, Fernando Gomes de; Silva, Fabrício Machado; Oliveira, Geiza Esperandio de

    2017-01-01

    After the consolidation of the Brazilian biodiesel industry, issues related to the final destination of the glycerin, the by-product from the biodiesel industrial process, drawing the attention of several researchers. There are several uses to this byproduct. Among them, the obtaining of polymers, such as polyurethane (PU), are very encouraged since the glycerin ca be used, as well as the castor oil, in the replacement of petrochemical polyols. The aim of this work was to propose a new route for the obtainment of a petroleum sorbent based on polyurethane resin from glycerin and castor oil, through the emulsion technique. In addition, maghemite (γ-Fe 2 O 3 ) was mixed to the polymer matrix, producing a magnetic composite, able to make easier the oil cleanup process. The products synthesized were characterized by Fourier transform infrared spectroscopy, X-ray diffraction, simultaneous Thermogravimetry (TGA) and Differential scanning calorimetry (DSC), Optical microscopy, Scanning electron microscopy (SEM). In addition, magnetic force and oil removal capability tests were also performed. The magnetic material was used to remove oil from water, exhibited a good oil removal capability. In a typical test, 1g of the composite containing 5wt% of maghemite was able to remove 10g of oil from water. (author)

  14. The influence of heat treatments on several types of base-metal removable partial denture alloys.

    Science.gov (United States)

    Morris, H F; Asgar, K; Rowe, A P; Nasjleti, C E

    1979-04-01

    Four removable partial denture alloys, Vitallium (Co-Cr alloy), Dentillium P.D. (Fe-Cr alloy), Durallium L.G. (Co-Cr-Ni alloy), and Ticonium 100 (Ni-Cr alloy), were evaluated in the as-cast condition and after heat treatment for 15 minutes at 1,300 degrees, 1,600 degrees, 1,900 degrees, and 2,200 degrees F followed by quenching in water. The following properties were determined and compared for each alloy at each heat treatment condition: the yield strengths at 0.01%, 0.1%, and 0.2% offsets, the ultimate tensile strength, the percent elongation, the modulus of elasticity, and the Knoop microhardness. The results were statistically analyzed. Photomicrographs were examined for each alloy and test condition. The following conclusions were made: 1. The "highest values" were exhibited by the as-cast alloy. 2. Heat treatment of the partial denture alloys tested resulted in reductions in strength, while the elongations varied. This study demonstrates that, in practice, one should avoid (a) prolonged "heat-soaking" while soldering and (b) grinding or polishing of the casting until the alloy is "red hot". 3. Durallium L.G. was the least affected by the various heat treatment conditions. 4. Conventional reporting of the yield strength at 0.2% offset, the ultimate tensile strength, and percent elongation are not adequate to completely describe and compare the mechanical behavior of alloys. The reporting of the yield strength at 0.01% offset, in addition to the other reported properties, will provide a more complete description of the behavior of the dental alloys.

  15. The effect of different aspect ratio and bottom heat flux towards contaminant removal using numerical analysis

    International Nuclear Information System (INIS)

    Saadun, M N A; Manaf, M Z A; Zakaria, M S; Hafidzal, M H M; Azwadi, C S Nor; Malek, Z A A

    2013-01-01

    Cubic Interpolated Pseudo-particle (CIP) numerical simulation scheme has been anticipated to predict the interaction involving fluids and solid particles in an open channel with rectangular shaped cavity flow. The rectangular shaped cavity is looking by different aspect ratio in modelling the real pipeline joints that are in a range of sizes. Various inlet velocities are also being applied in predicting various fluid flow characteristics. In this paper, the constant heat flux is introduced at the bottom wall, showing the buoyancy effects towards the contaminant's removal rate. In order to characterize the fluid flow, the numerical scheme alone is initially tested and validated in a lid driven cavity with a single particle. The study of buoyancy effects and different aspect ratio of rectangular geometry were carried out using a MATLAB govern by Navier-Stokes equation. CIP is used as a model for a numerical scheme solver for fluid solid particles interaction. The result shows that the higher aspect ratio coupled with heated bottom wall give higher percentage of contaminant's removal rate. Comparing with the benchmark results has demonstrated the applicability of the method to reproduce fluid structure which is complex in the system. Despite a slight deviation of the formations of vortices from some of the literature results, the general pattern is considered to be in close agreement with those published in the literature

  16. Design Report for the ½ Scale Air-Cooled RCCS Tests in the Natural convection Shutdown heat removal Test Facility (NSTF)

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Kilsdonk, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bremer, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Aeschlimann, R. W. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-06-01

    The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility was constructed in order to carry out highly instrumented experiments that can be used to validate the performance of passive safety systems for advanced reactor designs. The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems. Standing 25-m in height, the facility is able to supply up to 220 kW at 21 kW/m2 to accurately simulate the heat fluxes at the walls of a reactor pressure vessel. A suite of nearly 400 data acquisition channels, including a sophisticated fiber optic system for high density temperature measurements, guides test operations and provides data to support scaling analysis and modeling efforts. Measurements of system mass flow rate, air and surface temperatures, heat flux, humidity, and pressure differentials, among others; are part of this total generated data set. The following report provides an introduction to the top level-objectives of the program related to passively safe decay heat removal, a detailed description of the engineering specifications, design features, and dimensions of the test facility at Argonne. Specifications of the sensors and their placement on the test facility will be provided, along with a complete channel listing of the data acquisition system.

  17. A New Technique of Removing Blind Spots to Optimize Wireless Coverage in Indoor Area

    Directory of Open Access Journals (Sweden)

    A. W. Reza

    2013-01-01

    Full Text Available Blind spots (or bad sampling points in indoor areas are the positions where no signal exists (or the signal is too weak and the existence of a receiver within the blind spot decelerates the performance of the communication system. Therefore, it is one of the fundamental requirements to eliminate the blind spots from the indoor area and obtain the maximum coverage while designing the wireless networks. In this regard, this paper combines ray-tracing (RT, genetic algorithm (GA, depth first search (DFS, and branch-and-bound method as a new technique that guarantees the removal of blind spots and subsequently determines the optimal wireless coverage using minimum number of transmitters. The proposed system outperforms the existing techniques in terms of algorithmic complexity and demonstrates that the computation time can be reduced as high as 99% and 75%, respectively, as compared to existing algorithms. Moreover, in terms of experimental analysis, the coverage prediction successfully reaches 99% and, thus, the proposed coverage model effectively guarantees the removal of blind spots.

  18. BLINDAGE: A neutron and gamma-ray transport code for shieldings with the removal-diffusion technique coupled with the point-kernel technique

    International Nuclear Information System (INIS)

    Fanaro, L.C.C.B.

    1984-01-01

    It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt

  19. Sana experiments for self-acting removal of the after-heat in reactors with pebble bed fuel and their interpretation

    International Nuclear Information System (INIS)

    Niessen, H.F.; Stoecker, Bernd; Amoignon, Olivier; Zuying, Gao; Jie, Liu

    1997-01-01

    For the confirmation of self-acting afterheat removal under hypothetical accident conditions from pebble bed reactors at the Research Center Juelich a test facility with an electrical heating input up to 30kW was erected and operated. A description of the test facility is given. Within the different tests the pebble diameter, the pebble material, the gas in the pebble bed, the heating-power and the arrangement of the heating were changed. Parts of the data were used within an IAEA Co-ordinated Research Program as benchmark problems for the code validation. All computer codes could simulate the test results with a sufficient good agreement, when the tests were executed with helium. For the tests with nitrogen the natural convection has to be taken into account. (author)

  20. Verification of heat removal capability of a concrete cask system for spent fuel storage

    International Nuclear Information System (INIS)

    Sakai, Mikio; Fujiwara, Hiroaki; Sakaya, Tadatugu

    2001-01-01

    The reprocessing works comprising of a center of nuclear fuel cycle in Japan is now under construction at Rokkasho-mura in Aomori prefecture, which is to be operated in 2005. However, as reprocessing capacity of the works is under total forming amount of spent nuclear fuels, it has been essential to construct a new facility intermediately to store them at a period before reprocessing them because of prediction to reach limit of pool storage in nuclear power stations. There are some intermediate storage methods, which are water pool method for wet storage, and bolt method, metal cask method, silo method and concrete cask method for dry storage. Among many methods, the dry storage is focussed at a standpoint of its operability and economy, the concrete cask method which has a lot of using results in U.S.A. has been focussed as a method expectable in its cost reduction effect among it. The Ishikawajima-Harima Heavy Industries Co., Ltd. produced, in trial, a concrete cask with real size to confirm productivity when advancing design work on concrete cask. By using the trial product, a heat removal test mainly focussing temperature of concrete in the cask was carried out to confirm heat conductive performances of the cask. And, analysis of heat conductivity was also carried out to verify validity of its analysis model. (G.K.)

  1. Development of in-situ laser cutting technique for removal of single selected coolant channel from pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Vishwakarma, S.C.; Upadhyaya, B.N.

    2016-01-01

    We report on the development of a pulsed Nd:YAG laser based cutting technique for removal of single coolant channel from pressurized heavy water reactor (PHWR). It includes development of special tools/manipulators and optimization of laser cutting process parameters for cutting of liner tube, end fitting, bellow lip weld joint, and pressure tube stubs. For each cutting operation, a special tool with precision motion control is utilized. These manipulators/tools hold and move the laser cutting nozzle in the required manner and are fixed on the same coolant channel, which has to be removed. This laser cutting technique has been successfully deployed for removal of selected coolant channels Q-16, Q-15 and N-6 of KAPS-2 reactor with minimum radiation dose consumption and in short time. (author)

  2. Efficacy of the World Health Organization-recommended handwashing technique and a modified washing technique to remove Clostridium difficile from hands.

    Science.gov (United States)

    Deschênes, Philippe; Chano, Frédéric; Dionne, Léa-Laurence; Pittet, Didier; Longtin, Yves

    2017-08-01

    The efficacy of the World Health Organization (WHO)-recommended handwashing technique against Clostridium difficile is uncertain, and whether it could be improved remains unknown. Also, the benefit of using a structured technique instead of an unstructured technique remains unclear. This study was a prospective comparison of 3 techniques (unstructured, WHO, and a novel technique dubbed WHO shortened repeated [WHO-SR] technique) to remove C difficile. Ten participants were enrolled and performed each technique. Hands were contaminated with 3 × 10 6 colony forming units (CFU) of a nontoxigenic strain containing 90% spores. Efficacy was assessed using the whole-hand method. The relative efficacy of each technique and of a structured (either WHO or WHO-SR) vs an unstructured technique were assessed by Mann-Whitney U test and Wilcoxon signed-rank test. The median effectiveness of the unstructured, WHO, and WHO-SR techniques in log 10 CFU reduction was 1.30 (interquartile range [IQR], 1.27-1.43), 1.71 (IQR, 1.34-1.91), and 1.70 (IQR, 1.54-2.42), respectively. The WHO-SR technique was significantly more efficacious than the unstructured technique (P = .01). Washing hands with a structured technique was more effective than washing with an unstructured technique (median, 1.70 vs 1.30 log 10 CFU reduction, respectively; P = .007). A structured washing technique is more effective than an unstructured technique against C difficile. Copyright © 2017 Association for Professionals in Infection Control and Epidemiology, Inc. Published by Elsevier Inc. All rights reserved.

  3. Study on Natural Convection around a vertical heated rod using PIV/LIF technique,

    International Nuclear Information System (INIS)

    Szijarto, R.; Yamaji, B.; Aszodi, A.

    2010-01-01

    The Nuclear Training Reactor of the Institute of Nuclear Techniques (Budapest University of Technology and Economics, Hungary) is a pool-type reactor with light water moderator and with a maximum thermal power of 100 kW. The fuel elements are cooled by natural convection. An experimental setup was built to analyse the nature of the natural convection around a heated rod. The flow field was investigated using an electrically heated rod, which models the geometry of a fuel pin in the training reactor. The electric power of the model rod is variable between 0-500 W. The rod was placed in a square-based glass tank. Particle Image Velocimetry and Laser Induced Fluorescence measurement techniques were used to study the velocity and temperature field in a two-dimensional area. The thermal and the hydraulic boundary layers were detected near a rod in a lower section of the aquarium. The laminar-turbulent transition of the flow regime was observed, the maximum velocity of the up-flow was 0.025-0.05 m/s. From the temperature measurements the local heat transfer coefficient was estimated. (Authors)

  4. A new and simple extraction technique for rectal foreign bodies: removing by cutting into small pieces

    OpenAIRE

    Abbas Aras; Mehmet Karabulut; Osman Kones; Kaplan Baha Temizgonul; Halil Alis

    2014-01-01

    The purposes of insertion and types of foreign bodies in rectum show great variation. Rectal foreign bodies need to be removed without giving damage to intestinal wall and this should be done in the easiest possible way. We have reported a new and a simple technique. It is easy to apply and safe. A patient was admitted to our clinic with a rectal foreign body (radish) which was successfully removed by cutting it into small pieces. We conclude that different kinds of rectal foreign bodies, esp...

  5. Midline sclerotomy approach for intraocular foreign body removal in phakic eyes using endoilluminator: A novel technique

    Science.gov (United States)

    Ravani, Raghav; Chawla, Rohan; Azad, Shorya Vardhan; Gupta, Yogita; Kumar, Vinod; Kumar, Atul

    2018-01-01

    Purpose: The objective of this study is to describe the removal of retained intraocular foreign body (RIOFB) by bimanual pars plana vitrectomy through midline sclerotomy in phakic patients. Technique: Four eyes with RIOFB and clear lens underwent microincision vitrectomy surgery. A chandelier illumination was placed through one of the existing ports. The foreign body (FB) was localized by direct visualization (intravitreal) or indentation (pars plana), stabilized using an intraocular magnet/FB forceps introduced through a midline sclerotomy and freed of vitreous from all sides using a vitrectomy cutter through the other port bimanually, reoriented along their long axis and extracted through the midline sclerotomy. Results: All four FBs were removed successfully without slippage or damage to the clear lens. Conclusion: Chandelier illumination-assisted removal of FB through midline sclerotomy helps in easier localization, stabilization and removal, avoiding lens touch even in anteriorly located FBs such as at pars plana. PMID:29676316

  6. Economics of long distance transmission, storage and distribution of heat from nuclear plants with existing and newer techniques

    International Nuclear Information System (INIS)

    Margen, Peter

    1977-01-01

    Nuclear plants can provide heat for district heating in mainly two ways. Central nuclear power plants sufficiently large to be economic as electricity producers could instead be designed for heat extraction at temperatures useful for district heating. The second promising way is to design simple low temperature reactors, so simple and safe that near urban location becomes feasible. The manner of transport distribution and storage of heat is discussed in this paper which are very important especially in the cost calculations. The economic objectives can often be attained already with conventional technigues even when transport distances are large. But newer techniques of transport promise to make even cities at greater distances from major nuclear power plants economically connectible whilst new techniques for small distribution pipes help to extend the economic distribution area to the less dense one-family house districts. (M.S.)

  7. Technique for in-place welding of aluminum backed up by a combustible material

    Science.gov (United States)

    Spagnuolo, A. C.

    1971-01-01

    Welding external aluminum jacket, tightly wrapped around inner layer of wood composition fiberboard, in oxygen free environment prevents combustion and subsequent damage to underlying fiberboard. Technique also applies to metal cutting in similar assemblies without disassembly to remove combustible materials from welding heat proximity.

  8. Th effectiveness of soot removal techniques for the recovery of fingerprints on glass fire debris in petrol bomb cases

    International Nuclear Information System (INIS)

    Umi Kalthom Ahmad; Mei, Y.S.; Mohd Shahru Bahari; Raramasivam, V.K.

    2011-01-01

    The increased use of petrol bombs as an act of vengeance in Malaysia has heightened awareness for the need of research relating physical evidence found at the crime scene to the perpetrator of the crime. A study was therefore carried out to assess the effectiveness of soot removal techniques on glass fire debris without affecting the fingerprints found on the evidence. Soot was removed using three methods which were brushing, 2 % NaOH solution and tape lifting. Depending on the visibility of prints recovered, prints which were visible after soot removal were lifted directly while prints that were not visible were subjected to enhancement. Glass microscope slides were used in laboratory experiment and subjected to control burn for the formation of soot. Soot was later removed following enhancement of the prints over time (within 1 day, within 2 days and after 2 days). While in simulated petrol bomb ground experiment, petrol bombs were hurled in glass bottles and the fragments were collected. Favorable results were obtained in varying degrees using each soot removal methods. In laboratory testing, brushing and 2 % NaOH solution revealed fingerprints that were visible after removal of excess soot and were lifted directly. As for tape lifting technique, some prints were visible and were successfully lifted while those that were not visible were subjected to super glue fuming for effective fingerprint identification. (author)

  9. Heat pipes for ground heating and cooling

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, L L

    1988-01-01

    Different versions of heat pipe ground heating and cooling devices are considered. Solar energy, biomass, ground stored energy, recovered heat of industrial enterprises and ambient cold air are used as energy and cold sources. Heat pipe utilization of air in winter makes it possible to design accumulators of cold and ensures deep freezing of ground in order to increase its mechanical strength when building roadways through the swamps and ponds in Siberia. Long-term underground heat storage systems are considered, in which the solar and biomass energy is accumulated and then transferred to heat dwellings and greenhouses, as well as to remove snow from roadways with the help of heat pipes and solar collectors.

  10. A unified approach to assess performance of different techniques for recovering exhaust heat from gas turbines

    International Nuclear Information System (INIS)

    Carapellucci, Roberto

    2009-01-01

    Exhaust heat from gas turbines can be recovered externally or internally to the cycle itself. Of the technology options for external recovery, the combined gas-steam power plant is by far the most effective and commonly used worldwide. For internal recovery conventional solutions are based on thermodynamic regeneration and steam injection, while innovative solutions rely on humid air regeneration and steam reforming of fuel. In this paper a unified approach for analysing different exhaust heat recovery techniques is proposed. It has been possible to define a characteristic internal heat recovery plane, based on a few meaningful parameters and to identify an innovative scheme for repowering existing combined cycles. The characteristic plane indicates directly the performance obtainable with the different recovery techniques, showing that performances close to combined cycle plants (external recovery) can only be achieved with combined recovery techniques (humid air regeneration, steam reforming of fuel). The innovative repowering scheme, which requires the addition of a gas turbine and one-pressure level HRSG to an existing combined gas-steam power plant, significantly increases power output with fairly high marginal efficiency.

  11. Natural Circulation in the Blanket Heat Removal System During a Loss-of-Pumping Accident (LOFA) Based on Initial Conceptual Design

    International Nuclear Information System (INIS)

    Hamm, L.L.

    1998-01-01

    A transient natural convection model of the APT blanket primary heat removal (HR) system was developed to demonstrate that the blanket could be cooled for a sufficient period of time for long term cooling to be established following a loss-of-flow accident (LOFA). The particular case of interest in this report is a complete loss-of-pumping accident. For the accident scenario in which pumps are lost in both the target and blanket HR systems, natural convection provides effective cooling of the blanket for approximately 68 hours, and, if only the blanket HR systems are involved, natural convection is effective for approximately 210 hours. The heat sink for both of these accident scenarios is the assumed stagnant fluid and metal on the secondary sides of the heat exchangers

  12. Modeling of a heat sink and high heat flux vapor chamber

    Science.gov (United States)

    Vadnjal, Aleksander

    An increasing demand for a higher heat flux removal capability within a smaller volume for high power electronics led us to focus on a novel cold plate design. A high heat flux evaporator and micro channel heat sink are the main components of a cold plate which is capable of removing couple of 100 W/cm2. In order to describe performance of such porous media device a proper modeling has to be addressed. A universal approach based on the volume average theory (VAT) to transport phenomena in porous media is shown. An approach on how to treat the closure for momentum and energy equations is addressed and a proper definition for friction factors and heat transfer coefficients are discussed. A numerical scheme using a solution to Navier-Stokes equations over a representative elementary volume (REV) and the use of VAT is developed to show how to compute friction factors and heat transfer coefficients. The calculation show good agreement with the experimental data. For the heat transfer coefficient closure, a proper average for both fluid and solid is investigated. Different types of heating are also investigated in order to determine how it influences the heat transfer coefficient. A higher heat fluxes in small area condensers led us to the micro channels in contrast to the classical heat fin design. A micro channel can have various shapes to enhance heat transfer, but the shape that will lead to a higher heat flux removal with a moderate pumping power needs to be determined. The standard micro-channel terminology is usually used for channels with a simple cross section, e.g. square, round, triangle, etc., but here the micro channel cross section is going to be expanded to describe more complicated and interconnected micro scale channel cross sections. The micro channel geometries explored are pin fins (in-line and staggered) and sintered porous micro channels. The problem solved here is a conjugate problem involving two heat transfer mechanisms; (1) porous media

  13. Mercury and tritium removal from DOE waste oils

    Energy Technology Data Exchange (ETDEWEB)

    Klasson, E.T. [Oak Ridge National Lab., TN (United States)

    1997-10-01

    This work covers the investigation of vacuum extraction as a means to remove tritiated contamination as well as the removal via sorption of dissolved mercury from contaminated oils. The radiation damage in oils from tritium causes production of hydrogen, methane, and low-molecular-weight hydrocarbons. When tritium gas is present in the oil, the tritium atom is incorporated into the formed hydrocarbons. The transformer industry measures gas content/composition of transformer oils as a diagnostic tool for the transformers` condition. The analytical approach (ASTM D3612-90) used for these measurements is vacuum extraction of all gases (H{sub 2}, N{sub 2}, O{sub 2}, CO, CO{sub 2}, etc.) followed by analysis of the evolved gas mixture. This extraction method will be adapted to remove dissolved gases (including tritium) from the SRS vacuum pump oil. It may be necessary to heat (60{degrees}C to 70{degrees}C) the oil during vacuum extraction to remove tritiated water. A method described in the procedures is a stripper column extraction, in which a carrier gas (argon) is used to remove dissolved gases from oil that is dispersed on high surface area beads. This method appears promising for scale-up as a treatment process, and a modified process is also being used as a dewatering technique by SD Myers, Inc. (a transformer consulting company) for transformers in the field by a mobile unit. Although some mercury may be removed during the vacuum extraction, the most common technique for removing mercury from oil is by using sulfur-impregnated activated carbon (SIAC). SIAC is currently being used by the petroleum industry to remove mercury from hydrocarbon mixtures, but the sorbent has not been previously tested on DOE vacuum oil waste. It is anticipated that a final process will be similar to technologies used by the petroleum industry and is comparable to ion exchange operations in large column-type reactors.

  14. Comparison of the removal of calcium hydroxide medicaments on the root canal treatment irrigated with manual and sonic agitation technique

    Directory of Open Access Journals (Sweden)

    Anna Muryani

    2017-11-01

    Full Text Available Introduction: Irrigation of the root canal system is an important part of the endodontic treatment principle which aims to improve the hygiene of the root canal system from any debris and medicament residue with the hydrodynamic system. Root irrigation technique can be done with the manual and sonic system by using 2.5% NaOCI irrigation solution. Calcium hydroxide is used as a medicament for root canal sterilization. Root canal treatment will fail due to the imperfect removal of calcium hydroxide residue. The objective of this research was to analyze the comparison of the removal of calcium hydroxide medicaments on the root canal treatment irrigated with manual and sonic agitation technique using 2.5% sodium hypochlorite solution. Methods: The methods used in this study was experimental laboratory. The sample used was 30 maxillary incisors. The teeth were then divided into two groups randomly, then the root canal preparation was done by the crown down technique with manual irrigation using 2.5% NaOCI solution. The radicular part of the teeth was then split longitudinally, given a standardized groove in the one-third of the apical part, then applied with water-solved calcium hydroxide. The teeth were unified afterwards by using flowable composites, then soaked in the artificial saliva at the temperature of 37ºC. The sample of the 1st group was irrigated by manual agitation technique, and the sample of the 2nd group 2 was irrigated by sonic agitation technique, then both were viewed by stereo microscope. The data results were analyzed by Kruskal-Wallis and Mann-Whitney tests. Results: The results of calcium hydroxide removal were different between the root canals that were irrigated using 2.5% sodium hypochlorite irrigation solution by manual agitation technique compared to the sonic agitation technique. Irrigation using 2.5% sodium hypochlorite irrigation solution with the sonic agitation techniques were proven to be more effective in lifting Ca

  15. THE DIAGNOSIS TECHNIQUE OF ABNORMAL HEATING OF POWER CAPACITORS

    Directory of Open Access Journals (Sweden)

    D. I. Zalizny

    2016-01-01

    Full Text Available The existing system of protection and diagnostics are not able to detect abnormal heating of the power capacitors caused by its internal malfunction formation. The paper contains a proposal of a technique that enables to detect such heat at its early study. This technique consists of a hardware and an algorithms. The hardware consists of a microprocessor-based instrument developed by the author, of measuring transformers of current and of temperature sensors. This equipment must be connected to the condenser unit with a rated voltage of 380 V. In operation, the device performs continuous measurement of the surface temperature of the casing of each condenser, the temperature of the external environment, voltage and current from the power source. The measured values are used in the mathematical model of thermal processes that enables to calculate the temperature of the hottest point of each capacitor in real-time. Then the calculation of the intrinsic difference ΔΘ1° between the average temperature values of the dielectric and the base average value of this temperatures during the second day from the start of the measurements. If the Dq1 value exceeds the value of the absolute error of simulation, diagnostic signals of abnormal levels of heating, viz. low, medium, high and very high, are generated. It is also necessary to calculate the rate of change of ΔΘ1° and to consider the values obtained in the formation of hazard levels. For the low level and the average level of hazard the operation of diagnostic system with a visual signal is recommended, while for the high level of hazard it is recommended to use both visual and sound signals, and for the very high hazard level the capacitor ought to be turned off from the source. The algorithms have been developed heuristically. The final formation of the algorithms is possible only after the long-term operation of the proposed diagnosis system on real objects. The implementation of the

  16. Evaluation of Heat Transfer to the Implant-Bone Interface During Removal of Metal Copings Cemented onto Titanium Abutments.

    Science.gov (United States)

    Cakan, Umut; Cakan, Murat; Delilbasi, Cagri

    2016-01-01

    The aim of this investigation was to measure the temperature increase due to heat transferred to the implant-bone interface when the abutment screw channel is accessed or a metal-ceramic crown is sectioned buccally with diamond or tungsten carbide bur using an air rotor, with or without irrigation. Cobalt-chromium copings were cemented onto straight titanium abutments. The temperature changes during removal of the copings were recorded over a period of 1 minute. The sectioning of coping with diamond bur and without water irrigation generated the highest temperature change at the cervical part of the implant. Both crown removal methods resulted in an increase in temperature at the implant-bone interface. However, this temperature change did not exceed 47°C, the potentially damaging threshold for bone reported in the literature.

  17. The use of ferrofluids for heat removal: Advantage or disadvantage?

    Energy Technology Data Exchange (ETDEWEB)

    Krauzina, Marina T., E-mail: krauzina@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Bozhko, Aleksandra A., E-mail: bozhko@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Krauzin, Pavel V., E-mail: krauzin@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Suslov, Sergey A., E-mail: ssuslov@swin.edu.au [Department of Mathematics H38, Swinburne University of Technology, Hawthorn, Victoria 3122 (Australia)

    2017-06-01

    It is shown experimentally that, depending on the relative orientation of the gravity and the thermal gradient and on the pre-history of experiment, the application of a uniform external vertical magnetic field to a spherical cavity filled with magnetic ferrofluid can either enhance or suppress a convective heat transfer. - Highlights: • Conduction heat transfer in magnetic fluid heated from above is stronger than that in a fluid not containing nanoparticles. • The application of a uniform vertical magnetic field enhances heat transfer when magnetic fluid is heated from above. • Heat transfer in a magnetic fluid heated from below is weaker than that in a fluid not containing nanoparticles.

  18. Radon removal using point-of-entry water-treatment techniques. Final report, October 1988-June 1990

    International Nuclear Information System (INIS)

    Kinner, N.E.; Malley, J.P.; Clement, J.A.

    1990-10-01

    The purpose of the EPA Cooperative Agreement was to evaluate the performance of POE granular activated carbon (GAC), and diffused bubble and bubble place aeration systems treating a ground water supply containing radon (35,620 + or - 6,717 pCi/L). The pattern of loading to the units was designed to simulate daily demand in a household. Each of the systems was evaluated with respect to three primary factors: radon removal efficiency, potential problems, and economics. The radon removal efficiencies of the POE GAC units gradually deteriorated over time from 99.7% to 79% for the GAC without pretreatment and 99.7% to 85% for the units preceded by ion exchange. The bubble plate and diffused bubble POE units were very efficient (99%) at removing radon from the water. The resilience is primarly due to the high air to water ratios supplied by the aeration blowers. One major problem associated with the aeration techniques is iron oxidation/precipitation

  19. Tests of heat techniques in households. Analysis of the results of the field tests; Praktijkprestaties van warmtetechnieken bij huishoudens. Analyse resultaten veldtesten

    Energy Technology Data Exchange (ETDEWEB)

    De Jong, A.; Friedel, P.; Overman, P. [Energy Matters, Driebergen (Netherlands)

    2012-11-15

    The development within conventional techniques and new techniques for the recovery and generation of heat in the house construction industry led to a need for knowledge of efficiencies of those techniques in practice. Therefore a project has been set up to gain insight into the efficiencies. In the field tests, five heat techniques are investigated: high-efficiency boilers, solar water heaters, balanced ventilation systems with heat recovery (also called heat recovery systems) and heat pump water heaters [Dutch] De ontwikkeling binnen conventionele technieken en nieuwe technieken voor de terugwinning en opwekking van warmte in de woningbouw leidden ertoe dat er bij verschillende partijen in de keten een kennisbehoefte is ontstaan naar de rendementen van deze technieken in de praktijk. Daartoe heeft AgentschapNL een project opgezet om meer inzicht te verkrijgen in deze rendementen. In de veldtesten worden vijf warmtechnieken bekeken: HR-ketels, HRe-ketels, zonneboilers, gebalanceerde ventilatiesystemen met warmteterugwinning (verder WTW-systemen genoemd) en warmtepompboilers. Deze worden op minutenbasis bemeten.

  20. Prediction technique for minimum-heat-flux (MHF)- point condition of saturated pool boiling

    International Nuclear Information System (INIS)

    Nishio, Shigefumi

    1987-01-01

    The temperature-controlled hypothesis for the minimum-heat-flux (MHF)-point condition, in which the MHF-point temperature is regarded as the controlling factor and is expected to be independent of surface configuration and dimensions, is inductively investigated for saturated pool-boiling. In this paper such features of the MHF-point condition are experimentally proved first. Secondly, a correlation of the MHF-point temperature is developed for the effect of system pressure. Finally, a simple technique based on this correlation is presented to estimate the effects of surface configuration, dimensions and system pressure on the minimum heat flux. (author)

  1. Comparison of Low Concentration and High Concentration Arsenic Removal Techniques and Evaluation of Concentration of Arsenic in Ground Water: A Case Study of Lahore, Pakistan

    Energy Technology Data Exchange (ETDEWEB)

    Yasar, Abdullah; Tabinda, Amtul Bari; Shahzadi, Uzma; Saleem, Pakeeza [GC University, Lahore (Pakistan)

    2014-10-15

    The main focus of this study was the evaluation of arsenic concentration in the ground water of Lahore at different depth and application of different mitigation techniques for arsenic removal. Twenty four hours of solar oxidation gives 90% of arsenic removal as compared to 8 hr. or 16 hr. Among oxides, calcium oxide gives 96% of As removal as compared to 93% by lanthanum oxide. Arsenic removal efficiency was up to 97% by ferric chloride, whereas 95% by alum. Activated alumina showed 99% removal as compared to 97% and 95% removal with bauxite and charcoal, respectively. Elemental analysis of adsorbents showed that the presence of phosphate and silica can cause a reduction of arsenic removal efficiency by activated alumina, bauxite and charcoal. This study has laid a foundation for further research on arsenic in the city of Lahore and has also provided suitable techniques for arsenic removal.

  2. Comparison of Low Concentration and High Concentration Arsenic Removal Techniques and Evaluation of Concentration of Arsenic in Ground Water: A Case Study of Lahore, Pakistan

    International Nuclear Information System (INIS)

    Yasar, Abdullah; Tabinda, Amtul Bari; Shahzadi, Uzma; Saleem, Pakeeza

    2014-01-01

    The main focus of this study was the evaluation of arsenic concentration in the ground water of Lahore at different depth and application of different mitigation techniques for arsenic removal. Twenty four hours of solar oxidation gives 90% of arsenic removal as compared to 8 hr. or 16 hr. Among oxides, calcium oxide gives 96% of As removal as compared to 93% by lanthanum oxide. Arsenic removal efficiency was up to 97% by ferric chloride, whereas 95% by alum. Activated alumina showed 99% removal as compared to 97% and 95% removal with bauxite and charcoal, respectively. Elemental analysis of adsorbents showed that the presence of phosphate and silica can cause a reduction of arsenic removal efficiency by activated alumina, bauxite and charcoal. This study has laid a foundation for further research on arsenic in the city of Lahore and has also provided suitable techniques for arsenic removal

  3. Residual heat removal pump and low pressure safety injection pump retrofit program

    International Nuclear Information System (INIS)

    Dudiak, J.G.; McKenna, J.M.

    1992-01-01

    Residual Heat Removal (RHR) and low pressure safety injection (LPSI) pumps installed in pressurized water-to-reactor power plants are used to provide low-head safety injection in the event of loss of coolant in the reactor coolant system. Because these pumps are subjected to rather severe temperature and pressure transients, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. Typically the pump impeller is mounted on an extended motor shaft (close-coupled configuration) and a mechanical seal is employed at the pump end of the shaft. Traditionally RHR and LPSI pumps have been a significant maintenance item for many utilities. Periodic mechanical seal of motor bearing replacement often is considered routine maintenance. The closed-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. This paper introduces a design modification developed to convert the close-coupled RHR and LPSI pumps to a coupled configuration

  4. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Hung, T.C.; Dhir, V.K.; Chang, J.C.; Wang, S.K.

    2011-01-01

    Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 o C which is substantially lower than ∼627 o C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

  5. Thermal-hydraulic processes involved in loss of residual heat removal during reduced inventory operation

    International Nuclear Information System (INIS)

    Fletcher, C.D.; McHugh, P.R.; Naff, S.A.; Johnsen, G.W.

    1991-02-01

    This paper identifies the topics needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that would be used for each cooling mode, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain, core water boil-off, and reflux cooling processes. Important subcategories of the reflux cooling processes include: the initiation of reflux cooling from various plant conditions, the effects of air on reflux cooling, core level depression effects, issues regarding the steam generator secondaries, and the special case of boiler-condenser cooling with once-through steam generators. 25 refs., 6 figs., 1 tab

  6. Biodesulfurization techniques: Application of selected microorganisms for organic sulfur removal from coals. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Elmore, B.B.

    1993-08-01

    As an alternative to post-combustion desulfurization of coal and pre-combustion desulfurization using physicochemical techniques, the microbial desulfurization of coal may be accomplished through the use of microbial cultures that, in an application of various microbial species, may remove both the pyritic and organic fractions of sulfur found in coal. Organisms have been isolated that readily depyritize coal but often at prohibitively low rates of desulfurization. Microbes have also been isolated that may potentially remove the organic-sulfur fraction present in coal (showing promise when acting on organic sulfur model compounds such as dibenzothiophene). The isolation and study of microorganisms demonstrating a potential for removing organic sulfur from coal has been undertaken in this project. Additionally, the organisms and mechanisms by which coal is microbially depyritized has been investigated. Three cultures were isolated that grew on dibenzothiophene (DBT), a model organic-sulfur compound, as the sole sulfur source. These cultures (UMX3, UMX9, and IGTS8) also grew on coal samples as the sole sulfur source. Numerous techniques for pretreating and ``cotreating`` coal for depyritization were also evaluated for the ability to improve the rate or extent of microbial depyritization. These include prewashing the coal with various solvents and adding surfactants to the culture broth. Using a bituminous coal containing 0.61% (w/w) pyrite washed with organic solvents at low slurry concentrations (2% w/v), the extent of depyritization was increased approximately 25% in two weeks as compared to controls. At slurry concentrations of 20% w/v, a tetrachloroethylene treatment of the coal followed by depyritization with Thiobacillus ferrooxidans increased both the rate and extent of depyritization by approximately 10%.

  7. Preliminary review of critical shutdown heat removal items for common cause failure susceptibility on LMFBR's. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Allard, L.T.; Elerath, J.G.

    1976-02-01

    This document presents a common cause failure analysis for Critical LMFBR Shutdown Heat Removal Systems. The report is intended to outline a systematic approach to defining areas with significant potential for common causes of failure, and ultimately provide inputs to the reliability prediction model. A preliminary evaluation of postulatd single initiating causes resulting in multiple failures of LMFBR-SHRS items is presented in Appendix C. This document will be periodically updated to reflect new information and activity.

  8. Studies of decontamination using easy removable coatings technique

    International Nuclear Information System (INIS)

    Oglaza, J.; Nowak, Z.

    1991-01-01

    The usefulness of removable coatings for decontamination of steel and epoxy-resin painted surfaces was examined. Natural latex, Revultex, butadiene-styrene latex as well as mixtures of latex with complexing agents and surfactants were used as decontaminating coats. The best decontamination was obtained by removable coatings of natural latex with EDTA additive for all surfaces and radionuclides tested. (author). 16 refs, 5 tabs

  9. Experimental validation of energy parameters in parabolic trough collector with plain absorber and analysis of heat transfer enhancement techniques

    Science.gov (United States)

    Bilal, F. R.; Arunachala, U. C.; Sandeep, H. M.

    2018-01-01

    The quantum of heat loss from the receiver of the Parabolic Trough Collector is considerable which results in lower thermal efficiency of the system. Hence heat transfer augmentation is essential which can be attained by various techniques. An analytical model to evaluate the system with bare receiver performance was developed using MATLAB. The experimental validation of the model resulted in less than 5.5% error in exit temperature using both water and thermic oil as heat transfer fluid. Further, heat transfer enhancement techniques were incorporated in the model which included the use of twisted tape inserts, nanofluid, and a combination of both for further enhancement. It was observed that the use of evacuated glass cover in the existing setup would increase the useful heat gain up to 5.3%. Fe3O4/H2O nanofluid showed a maximum enhancement of 56% in the Nusselt number for the volume concentration of 0.6% at highest Reynolds number. Similarly, twisted tape turbulators (with twist ratio of 2) taken alone with water exhibited 59% improvement in Nusselt number. Combining both the heat transfer augmentation techniques at their best values revealed the Nusselt number enhancement up to 87%. It is concluded that, use of twisted tape with water is the best method for heat transfer augmentation since it gives the maximum effective thermal efficiency amongst all for the range of Re considered. The first section in your paper

  10. Two new techniques for the remote evaluation of reactor steel condition - microscopic removal and surface examination

    International Nuclear Information System (INIS)

    Clayton, R.

    Much reactor inspection work involves an assessment of the condition of structural steel. This paper reviews two different techniques which provide information for such an assessment. The first - micro-sample removal (for the measurement of surface oxide thickness and chemical composition) - requires contact with the steel surface, whereas the second - a 'teach and learn' photographic technique (in which a special photogrammatic camera is used to obtain high-quality close-up photographs, to assess surface condition and corrosion growth) can obtain surface information on inaccessible components. (author)

  11. Piezosurgery for the Lingual Split Technique in Lingual Positioned Impacted Mandibular Third Molar Removal

    Science.gov (United States)

    Ge, Jing; Yang, Chi; Zheng, Jiawei; Qian, Wentao

    2016-01-01

    Abstract The aim of this study was to evaluate the effect and safety of lingual split technique using piezosurgery for the extraction of lingual positioned impacted mandibular 3rd molars with the goal of proposing a more minimally invasive choice for this common surgery. Eighty-nine consecutive patients with 110 lingual positioned impacted mandibular 3rd molars requiring extraction were performed the lingual split technique using piezosurgery. One sagittal osteotomy line and 2 transverse osteotomy line were designed for lingual and occlusal bone removal. The success rate, operative time, postoperative outcome, and major complications (including nerve injury, mandible fracture, severe hematoma or edema, and severe pyogenic infection) were documented and analyzed. All impacted mandibular 3rd molars were successfully removed (110/110). The average time of operation was 14.6 minutes (ranged from 7 to 28 minutes). One hundred and seven extraction sites (97.3%) were primary healing. Pain, mouth opening, swelling, and PoSSe scores on postoperative 7-day were 0.34 ± 0.63, 3.88 ± 0.66(cm), 2.4 ± 0.2(cm), and 23.7 ± 5.9, respectively. There were 6 cases (5.5%) had lingual nerve disturbance and 3 cases (2.7%) developed inferior alveolar nerve impairment, and achieved full recovery within 2 months by neurotrophic drug treatment. Our study suggested piezosurgery for lingual split technique provided an effective way for the extraction of lingual positioned and deeply impacted mandibular 3rd molar. PMID:27015214

  12. Effectiveness of four different final irrigation activation techniques on smear layer removal in curved root canals : a scanning electron microscopy study.

    Directory of Open Access Journals (Sweden)

    Puneet Ahuja

    2014-02-01

    Full Text Available The aim of this study was to assess the efficacy of apical negative pressure (ANP, manual dynamic agitation (MDA, passive ultrasonic irrigation (PUI and needle irrigation (NI as final irrigation activation techniques for smear layer removal in curved root canals.Mesiobuccal root canals of 80 freshly extracted maxillary first molars with curvatures ranging between 25° and 35° were used. A glide path with #08-15 K files was established before cleaning and shaping with Mtwo rotary instruments (VDW, Munich, Germany up to size 35/0.04 taper. During instrumentation, 1 ml of 2.5% NaOCl was used at each change of file. Samples were divided into 4 equal groups (n=20 according to the final irrigation activation technique: group 1, apical negative pressure (ANP (EndoVac; group 2, manual dynamic agitation (MDA; group 3, passive ultrasonic irrigation (PUI; and group 4, needle irrigation (NI. Root canals were split longitudinally and subjected to scanning electron microscopy. The presence of smear layer at coronal, middle and apical levels was evaluated by superimposing 300-μm square grid over the obtained photomicrographs using a four-score scale with X1,000 magnification.Amongst all the groups tested, ANP showed the overall best smear layer removal efficacy (p < 0.05. Removal of smear layer was least effective with the NI technique.ANP (EndoVac system can be used as the final irrigation activation technique for effective smear layer removal in curved root canals.

  13. Highly efficient removal of perfluorooctanoic acid from aqueous solution by H2O2-enhanced electrocoagulation-electroflotation technique

    Directory of Open Access Journals (Sweden)

    Bo Yang

    2016-03-01

    Full Text Available Electrocoagulation (EC technique was used to investigate the removal performance of aqueous perfluorooctanoic acid (PFOA with relatively high concentration as simulating the wastewater from organic fluorine industry. A comparison was done with the similar amount of coagulant between EC and chemical coagulation process. PFOA removal obtained was higher with EC process, especially for Fe anode. Several factors were studied to optimize the EC process. At the optimal operating parameters including 37.5 mA/cm2 of current density, initial pH 3.77, and 180 rpm of mixing speed, 93% of PFOA could be removed with 100 mg/L of initial concentration after 90-min electrolysis. Furthermore, the remove efficiency could be obviously improved by H2O2 intermittent addition, which removed more than 99% of PFOA within 40-min EC. It could be attributed to that H2O2 facilitated the oxidative transformation from ferrous to ferric ion. In addition, the adsorptive removal of aqueous PFOA on Fe flocs during EC was also verified by fourier transform infrared spectra.

  14. Experience on sodium removal from various components

    Energy Technology Data Exchange (ETDEWEB)

    Kamei, M; Kanbe, M; Yagisawa, H; Sasaki, S; Kataoka, H; Fukada, T; Ishii, Y; Saito, R; Mimoto, Y [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

    1978-08-01

    Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

  15. Experience on sodium removal from various components

    International Nuclear Information System (INIS)

    Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.; Fukada, T.; Ishii, Y.; Saito, R.; Mimoto, Y.

    1978-01-01

    Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

  16. Experience on sodium removal from various components

    International Nuclear Information System (INIS)

    Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.

    1978-02-01

    Since 1970, OEC (O-arai Engineering Center) has been investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of ''JOYO'' and Dummy fuel assembly of ''JOYO'', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of ''JOYO'', a sector model of Sodium-to-Air cooler of ''JOYO'' and a proto-type Isolation valve of ''JOYO'' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental subassemblies, the Fuel Handling Machine of ''MONJU'' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of Sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a ''JOYO'' prototype pump by reinstalling it after sodium removal five times. (author)

  17. Review of current status of high flux heat transfer techniques. Volume I. Text + Appendix A

    International Nuclear Information System (INIS)

    Bauer, W.H.; Gordon, H.S.; Lackner, H.; Mettling, J.R.; Miller, J.E.

    1980-09-01

    The scope of this work comprised two tasks. The first was to review high heat flux technology with consideration given to heat transfer panel configuration, diagnostics techniques and coolant supply. The second task was to prepare a report describing the findings of the review, to recommend the technology offering the least uncertainty for scale-up for the MFTF-B requirement and to recommend any new or perceived requirements for R and D effort

  18. Review of current status of high flux heat transfer techniques. Volume I. Text + Appendix A

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, W.H.; Gordon, H.S.; Lackner, H.; Mettling, J.R.; Miller, J.E.

    1980-09-01

    The scope of this work comprised two tasks. The first was to review high heat flux technology with consideration given to heat transfer panel configuration, diagnostics techniques and coolant supply. The second task was to prepare a report describing the findings of the review, to recommend the technology offering the least uncertainty for scale-up for the MFTF-B requirement and to recommend any new or perceived requirements for R and D effort.

  19. SLAG CHARACTERIZATION AND REMOVAL USING PULSE DETONATION TECHNOLOGY DURING COAL GASIFICATION

    Energy Technology Data Exchange (ETDEWEB)

    DR. DANIEL MEI; DR. JIANREN ZHOU; DR. PAUL O. BINEY; DR. ZIAUL HUQUE

    1998-07-30

    Pulse detonation technology for the purpose of removing slag and fouling deposits in coal-fired utility power plant boilers offers great potential. Conventional slag removal methods including soot blowers and water lances have great difficulties in removing slags especially from the down stream areas of utility power plant boilers. The detonation wave technique, based on high impact velocity with sufficient energy and thermal shock on the slag deposited on gas contact surfaces offers a convenient, inexpensive, yet efficient and effective way to supplement existing slag removal methods. A slight increase in the boiler efficiency, due to more effective ash/deposit removal and corresponding reduction in plant maintenance downtime and increased heat transfer efficiency, will save millions of dollars in operational costs. Reductions in toxic emissions will also be accomplished due to reduction in coal usage. Detonation waves have been demonstrated experimentally to have exceptionally high shearing capability, important to the task of removing slag and fouling deposits. The experimental results describe the parametric study of the input parameters in removing the different types of slag and operating condition. The experimental results show that both the single and multi shot detonation waves have high potential in effectively removing slag deposit from boiler heat transfer surfaces. The results obtained are encouraging and satisfactory. A good indication has also been obtained from the agreement with the preliminary computational fluid dynamics analysis that the wave impacts are more effective in removing slag deposits from tube bundles rather than single tube. This report presents results obtained in effectively removing three different types of slag (economizer, reheater, and air-heater) t a distance of up to 20 cm from the exit of the detonation tube. The experimental results show that the softer slags can be removed more easily. Also closer the slag to the exit of

  20. Thermal control system. [removing waste heat from industrial process spacecraft

    Science.gov (United States)

    Hewitt, D. R. (Inventor)

    1983-01-01

    The temperature of an exothermic process plant carried aboard an Earth orbiting spacecraft is regulated using a number of curved radiator panels accurately positioned in a circular arrangement to form an open receptacle. A module containing the process is insertable into the receptacle. Heat exchangers having broad exterior surfaces extending axially above the circumference of the module fit within arcuate spacings between adjacent radiator panels. Banks of variable conductance heat pipes partially embedded within and thermally coupled to the radiator panels extend across the spacings and are thermally coupled to broad exterior surfaces of the heat exchangers by flanges. Temperature sensors monitor the temperature of process fluid flowing from the module through the heat exchanges. Thermal conduction between the heat exchangers and the radiator panels is regulated by heating a control fluid within the heat pipes to vary the effective thermal length of the heat pipes in inverse proportion to changes in the temperature of the process fluid.

  1. Safety technology qualification of the prestressed cast iron pressure vessel (PCIV) and of the primary cell of the HTR-modul for the passive removal of decay heat, phase 1 (INHR)

    International Nuclear Information System (INIS)

    Warnke, E.P.

    1990-02-01

    During this development program the thermodynamic behaviour of a system was investigated, consisting of a hot working Prestressed Cast Iron Pressure Vessel and an inactive heat sink in the surrounding cavern cell. It could be shown, that the inactive heat removal system designed as a natural circuit can remove the maximum amount of heat of 890 kW during emergency conditions via a natural-draught air cooling tower even under very conservative assumptions and for a 50% loss of cooling pipes. Further it could be shown, that the hot working Prestressed Cast Iron Pressure Vessel has a very safe load carrying behaviour during all normal and upset conditions. (orig.) With 10 tabs., 38 figs., 43 refs [de

  2. Using geophysical techniques to control in situ thermal remediation

    International Nuclear Information System (INIS)

    Boyd, S.; Daily, W.; Ramirez, A.; Wilt, M.; Goldman, R.; Kayes, D.; Kenneally, K.; Udell, K.; Hunter, R.

    1994-01-01

    Monitoring the thermal and hydrologic processes that occur during thermal environmental remediation programs in near real-time provides essential information for controlling the process. Geophysical techniques played a crucial role in process control as well as for characterization during the recent Dynamic Underground Stripping Project demonstration in which several thousand gallons of gasoline were removed from heterogeneous soils both above and below the water table. Dynamic Underground Stripping combines steam injection and electrical heating for thermal enhancement with ground water pumping and vacuum extraction for contaminant removal. These processes produce rapid changes in the subsurface properties including changes in temperature fluid saturation, pressure and chemistry. Subsurface imaging methods are used to map the heated zones and control the thermal process. Temperature measurements made in wells throughout the field reveal details of the complex heating phenomena. Electrical resistance tomography (ERT) provides near real-time detailed images of the heated zones between boreholes both during electrical heating and steam injection. Borehole induction logs show close correlation with lithostratigraphy and, by identifying the more permeable gravel zones, can be used to predict steam movement. They are also useful in understanding the physical changes in the field and in interpreting the ERT images. Tiltmeters provide additional information regarding the shape of the steamed zones in plan view. They were used to track the growth of the steam front from individual injectors

  3. Recycling of rare earth magnet scraps: Carbon and oxygen removal from Nd magnet scraps

    International Nuclear Information System (INIS)

    Saguchi, A.; Asabe, K.; Fukuda, T.; Takahashi, W.; Suzuki, R.O.

    2006-01-01

    The decarburization and deoxidation technique for permanent Nd-Fe-B magnet scrap is investigated. The carbon and oxygen contamination damage the magnetic properties. The carbon content decreased less than 0.001% by heating in air. The two stage deoxidation is applied, iron oxides are reduced by heating in hydrogen thereafter rare earth oxides are removed by Ca-reduction and leaching. The appropriate conditions for deoxidation in the Ca-reduction and suppressing the re-oxidation in the leaching are investigated. The heating pattern in Ca-reduction and the leaching condition for the mixture composed of Ca compounds and Nd-Fe-B alloy powder greatly affects the oxygen content of recycled material. The decarburized and deoxidized Nd-Fe-B magnet scrap can be recycled as alloying elements by melting

  4. Production of molten UO2 pools by internal heating: apparatus and preliminary experimental heat transfer results

    International Nuclear Information System (INIS)

    Chasanov, M.G.; Gunther, W.H.; Baker, L. Jr.

    1977-01-01

    The capability for removal of heat from a pool of molten fuel under postaccident conditions is an important consideration in liquid-metal fast breeder reactor safety analysis. No experimental data for pool heat transfer from molten UO 2 under conditions simulating internal heat generation by fission product decay have been reported previously in the literature. An apparatus to provide such data was developed and used to investigate heat transfer from pools containing up to 7.5 kg of UO 2 ; the internal heat generation rates and pool depths attained cover most of the ranges of interest for postaccident heat removal analysis. It was also observed in these studies that the presence of simulated fission products corresponding to approximately 150,000 kW-day/kg burnup had no significant effect on the observed heat transfer

  5. Novel Power Electronics Three-Dimensional Heat Exchanger: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Bennion, K.; Cousineau, J.; Lustbader, J.; Narumanchi, S.

    2014-08-01

    Electric drive systems for vehicle propulsion enable technologies critical to meeting challenges for energy, environmental, and economic security. Enabling cost-effective electric drive systems requires reductions in inverter power semiconductor area. As critical components of the electric drive system are made smaller, heat removal becomes an increasing challenge. In this paper, we demonstrate an integrated approach to the design of thermal management systems for power semiconductors that matches the passive thermal resistance of the packaging with the active convective cooling performance of the heat exchanger. The heat exchanger concept builds on existing semiconductor thermal management improvements described in literature and patents, which include improved bonded interface materials, direct cooling of the semiconductor packages, and double-sided cooling. The key difference in the described concept is the achievement of high heat transfer performance with less aggressive cooling techniques by optimizing the passive and active heat transfer paths. An extruded aluminum design was selected because of its lower tooling cost, higher performance, and scalability in comparison to cast aluminum. Results demonstrated a heat flux improvement of a factor of two, and a package heat density improvement over 30%, which achieved the thermal performance targets.

  6. Taurolidine as an effective and biocompatible additive for plaque-removing techniques on implant surfaces.

    Science.gov (United States)

    John, Gordon; Schwarz, Frank; Becker, Jürgen

    2015-06-01

    The aim of the present study was the evaluation of the effectiveness and efficiency of two plaque-removing techniques, plastic curettes (PC) and glycine powder airflow (GLY) in combination with taurolidine (T), chlorhexidine (CHX), or pure water (PW) as additives and compared to groups without previous treatment (NT). Plaque was collected on titanium samples for 48 h in six subjects. Specimens were worn in a special splint in the upper jaw and randomly assigned to test and control groups. After biofilm removal procedures, clean implant surface (CIS) on the samples and treatment time were taken as parameters. Mean CIS was determined in the following descending order: T-GLY > CHX-GLY > NT-GLY > T-PC > PW-GLY > PW-PC > CHX-PC > NT-PC. Mean treatment time was determined in the following ascending order: T-GLY treatment times of the T groups were significantly lower than their corresponding PC or GLY groups. The results of the current study indicate that taurolidine seems to enhance effectiveness of plaque-removing procedures with plastic curettes and glycine powder airflow. Also, the efficiency of both treatment procedures seems to be increased.

  7. The Intra Uterine Morcellator: a new hysteroscopic operating technique to remove intrauterine polyps and myomas.

    Science.gov (United States)

    Emanuel, Mark Hans; Wamsteker, Kees

    2005-01-01

    A new hysteroscopic operating technique was compared retrospectively with conventional resectoscopy. Retrospective comparison (Canadian Task Force Classification II-2). Gynecology department of a university-affiliated teaching hospital. Fifty-five women, 27 with endometrial polyps and 28 with submucous myomas. Patients were treated with a prototype of the Intra Uterine Morcellator (IUM). This cutting device, 35 cm in length, was inserted into a straight working channel of a 90-mm hysteroscope. The major advantages were ease of removal of tissue fragments through the instrument and the use of saline solution instead of electrolyte-free solutions used in monopolar high-frequency resectoscopy. The mean operating time was 8.7 minutes (95% CI: 7.3-10.1) for the removal of endometrial polyps compared with 30.9 minutes (CI: 27.0-34.8) for resectoscopy, and 16.4 minutes (CI: 12.6-20.2) for submucous myomas compared with 42.2 minutes (CI: 39.7-44.7) for resectoscopy. All procedures were uneventful. This new technique is faster, and it appears to be easier to perform. Therefore, it can be expected to result in fewer fluid-related complications and to lead to a shorter learning curve when compared with conventional resectoscopy.

  8. Prehospital emergency removal of football helmets using two techniques.

    Science.gov (United States)

    Swartz, Erik E; Hernandez, Adam E; Decoster, Laura C; Mihalik, Jason P; Burns, Matthew F; Reynolds, Cathryn

    2011-01-01

    To compare the Eject Helmet Removal (EHR) System with manual football helmet removal. This quasiexperimental counterbalanced study was conducted in a controlled laboratory setting. Thirty certified athletic trainers (17 men and 13 women; mean ± standard deviation age: 33.03 ± 10.02 years; height: 174.53 ± 12.04 cm; mass: 85.19 ± 19.84 kg) participated after providing informed consent. Participants removed a Riddell Revolution IQ football helmet from a healthy model two times each under two conditions: manual helmet removal (MHR) and removal with the EHR system. A six-camera, three-dimensional motion capture system was used to record range of motion (ROM) of the head. A digital stopwatch was used to time trials and to record a split time associated with EHR system bladder insertion. A modified Borg CR10 scale was used to measure the rating of perceived exertion (RPE). Mean values were created for each variable. Three pairwise t-tests with Bonferroni-corrected alpha levels tested for differences between time for removal, split time, and RPE. A 2 x 3 (condition x plane) totally within-subjects repeated-measures design analysis of variance (ANOVA) tested for differences in head ROM between the sagittal, frontal, and transverse planes. Analyses were performed using SPSS (version 18.0) (alpha = 0.05). There was no statistically significant difference in perceived difficulty between EHR (RPE = 2.73) and MHR (RPE = 2.55) (t(29) = 0.76; p = 0.45; d = 0.20). Manual helmet removal was, on average, 28.95 seconds faster than EHR (t(29) = 11.44; p football helmets and in helmets used in other sports such as lacrosse, motorsports, and ice hockey.

  9. Thermal Conductivity of EB-PVD Thermal Barrier Coatings Evaluated by a Steady-State Laser Heat Flux Technique

    Science.gov (United States)

    Zhu, Dongming; Miller, Robert A.; Nagaraj, Ben A.; Bruce, Robert W.

    2000-01-01

    The thermal conductivity of electron beam-physical vapor deposited (EB-PVD) Zr02-8wt%Y2O3 thermal barrier coatings was determined by a steady-state heat flux laser technique. Thermal conductivity change kinetics of the EB-PVD ceramic coatings were also obtained in real time, at high temperatures, under the laser high heat flux, long term test conditions. The thermal conductivity increase due to micro-pore sintering and the decrease due to coating micro-delaminations in the EB-PVD coatings were evaluated for grooved and non-grooved EB-PVD coating systems under isothermal and thermal cycling conditions. The coating failure modes under the high heat flux test conditions were also investigated. The test technique provides a viable means for obtaining coating thermal conductivity data for use in design, development, and life prediction for engine applications.

  10. Molecular dynamics study of solid-liquid heat transfer and passive liquid flow

    Science.gov (United States)

    Yesudasan Daisy, Sumith

    High heat flux removal is a challenging problem in boilers, electronics cooling, concentrated photovoltaic and other power conversion devices. Heat transfer by phase change is one of the most efficient mechanisms for removing heat from a solid surface. Futuristic electronic devices are expected to generate more than 1000 W/cm2 of heat. Despite the advancements in microscale and nanoscale manufacturing, the maximum passive heat flux removal has been 300 W/cm2 in pool boiling. Such limitations can be overcome by developing nanoscale thin-film evaporation based devices, which however require a better understanding of surface interactions and liquid vapor phase change process. Evaporation based passive flow is an inspiration from the transpiration process that happens in trees. If we can mimic this process and develop heat removal devices, then we can develop efficient cooling devices. The existing passive flow based cooling devices still needs improvement to meet the future demands. To improve the efficiency and capacity of these devices, we need to explore and quantify the passive flow happening at nanoscales. Experimental techniques have not advanced enough to study these fundamental phenomena at the nanoscale, an alternative method is to perform theoretical study at nanoscales. Molecular dynamics (MD) simulation is a widely accepted powerful tool for studying a range of fundamental and engineering problems. MD simulations can be utilized to study the passive flow mechanism and heat transfer due to it. To study passive flow using MD, apart from the conventional methods available in MD, we need to have methods to simulate the heat transfer between solid and liquid, local pressure, surface tension, density, temperature calculation methods, realistic boundary conditions, etc. Heat transfer between solid and fluids has been a challenging area in MD simulations, and has only been minimally explored (especially for a practical fluid like water). Conventionally, an

  11. LMFBR post accident heat removal testing needs and conceptual design of a test facility

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

    1977-03-01

    A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

  12. Reliability analysis of 2400 MWth gas-cooled fast reactor natural circulation decay heat removal system

    International Nuclear Information System (INIS)

    Marques, M.; Bassi, C.; Bentivoglio, F.

    2012-01-01

    In support to a PSA (Probability Safety Assessment) performed at the design level on the 2400 MWth Gas-cooled Fast Reactor, the functional reliability of the decay heat removal system (DHR) working in natural circulation has been estimated in two transient situations corresponding to an 'aggravated' Loss of Flow Accident (LOFA) and a Loss of Coolant Accident (LOCA). The reliability analysis was based on the RMPS methodology. Reliability and global sensitivity analyses use uncertainty propagation by Monte Carlo techniques. The DHR system consists of 1) 3 dedicated DHR loops: the choice of 3 loops (3*100% redundancy) is made in assuming that one could be lost due to the accident initiating event (break for example) and that another one must be supposed unavailable (single failure criterion); 2) a metallic guard containment enclosing the primary system (referred as close containment), not pressurized in normal operation, having a free volume such as the fast primary helium expansion gives an equilibrium pressure of 1.0 MPa, in the first part of the transient (few hours). Each dedicated DHR loop designed to work in forced circulation with blowers or in natural circulation, is composed of 1) a primary loop (cross-duct connected to the core vessel), with a driving height of 10 meters between core and DHX mid-plan; 2) a secondary circuit filled with pressurized water at 1.0 MPa (driving height of 5 meters for natural circulation DHR); 3) a ternary pool, initially at 50 C. degrees, whose volume is determined to handle one day heat extraction (after this time delay, additional measures are foreseen to fill up the pool). The results obtained on the reliability of the DHR system and on the most important input parameters are very different from one scenario to the other showing the necessity for the PSA to perform specific reliability analysis of the passive system for each considered scenario. The analysis shows that the DHR system working in natural circulation is

  13. Study on heat removal capability concrete cask system with horizontal orientation

    International Nuclear Information System (INIS)

    Nabemoto, Toyonobu; Sakai, Mikio; Fujiwara, Hiroaki; Sakaya, Tadatsugu

    2002-01-01

    In Japan, nuclear fuel cycle, has been promoted, so the recycle fuels formed at nuclear power stations are planned to be processed at reprocessing facilities in future. However, as forming quantities of the recycle fuels are more than reprocessing quantities of the facilities, it is needed to practice a facility (interim storage facility (ISF)) to temporarily store them among the recycle fuels will be reprocessed. The Ishikawajima-Harima Heavy Industries, Co., Ltd. has investigated on vault system and concrete cask system for dry storage system with excellent economical efficiency among various systems on ISFs. As the latter method has a number of actual results in U.S.A., its practice is progressed after some improvements suitable for Japan. When progressing this practice on the latter method on fiscal year 1999, at first, a concrete cask with actual size was experimentally produced, to confirm its productivity. On fiscal year 2000, aiming to establish heat removal evaluation at storage, a thermal load test simulated at the storage was carried out by using this trial product. Here was reported results obtained at a test simulated at repacking carried out on fiscal year 2001. (G.K.)

  14. Removal of radon by aeration testing of various aeration techniques for small water works. For European Commission under Contract No FI4PCT960054 TENAWA project

    CERN Document Server

    Salonen, L; Mehtonen, J; Mjoenes, L; Raff, O; Turunen, H

    2002-01-01

    Capability of various aeration techniques to remove radon from water in small waterworks was studied as a part of project (Treatment Techniques for Removing Natural Radionuclides from Drinking Water), which was carried out during 1997-1999 on a cost-shared basis (contract No. F14PCT960054) with The European Commission (CEC) under the supervision of the Directorate-General XII Radiation Protection Research Unit. In TENAWA project both laboratory and field experiments were performed in order to find reliable methods and equipment for removing natural radionuclides from ground water originating either from private wells or small waterworks. Because such techniques are more often needed in private households than at waterworks, the main emphasis of the research was aimed to solve the water treatment problems related to the private water supplies, especially bedrock wells. Radon was the most important radionuclide to be removed from water at waterworks whereas the removal of other radionuclides ( sup 2 sup 3 sup 4...

  15. Development of a steady-state calculation model for the KALIMER PDRC(Passive Decay Heat Removal Circuit)

    International Nuclear Information System (INIS)

    Chang, Won Pyo; Ha, Kwi Seok; Jeong, Hae Yong; Kwon, Young Min; Eoh, Jae Hyuk; Lee, Yong Bum

    2003-06-01

    A sodium circuit has usually featured for a Liquid Metal Reactor(LMR) using sodium as coolant to remove the decay heat ultimately under accidental conditions because of its high reliability. Most of the system codes used for a Light Water Reactor(LWR) analysis is capable of calculating natural circulation within such circuit, but the code currently used for the LMR analysis does not feature stand alone capability to simulate the natural circulation flow inside the circuit due to its application limitation. To this end, the present study has been carried out because the natural circulation analysis for such the circuit is realistically raised for the design with a new concept. The steady state modeling is presented in this paper, development of a transient model is also followed to close the study. The incompressibility assumption of sodium which allow the circuit to be modeled with a single flow, makes the model greatly simplified. Models such as a heat exchanger developed in the study can be effectively applied to other system analysis codes which require such component models

  16. Chemical heat pump and chemical energy storage system

    Science.gov (United States)

    Clark, Edward C.; Huxtable, Douglas D.

    1985-08-06

    A chemical heat pump and storage system employs sulfuric acid and water. In one form, the system includes a generator and condenser, an evaporator and absorber, aqueous acid solution storage and water storage. During a charging cycle, heat is provided to the generator from a heat source to concentrate the acid solution while heat is removed from the condenser to condense the water vapor produced in the generator. Water is then stored in the storage tank. Heat is thus stored in the form of chemical energy in the concentrated acid. The heat removed from the water vapor can be supplied to a heat load of proper temperature or can be rejected. During a discharge cycle, water in the evaporator is supplied with heat to generate water vapor, which is transmitted to the absorber where it is condensed and absorbed into the concentrated acid. Both heats of dilution and condensation of water are removed from the thus diluted acid. During the discharge cycle the system functions as a heat pump in which heat is added to the system at a low temperature and removed from the system at a high temperature. The diluted acid is stored in an acid storage tank or is routed directly to the generator for reconcentration. The generator, condenser, evaporator, and absorber all are operated under pressure conditions specified by the desired temperature levels for a given application. The storage tanks, however, can be maintained at or near ambient pressure conditions. In another form, the heat pump system is employed to provide usable heat from waste process heat by upgrading the temperature of the waste heat.

  17. Some techniques for sodium removal in CIAE

    International Nuclear Information System (INIS)

    Yuan Waimai; Ding Dejun; Guo Huanfang; Hong Shuzhang; Zhou Shuxia; Shen Fenyang; Yang Zhongmin; Xu Yongxing

    1997-01-01

    In this paper the experiment and application on sodium removal and sodium disposal are presented. Steam-nitrogen process was used in CIAE for cleaning cold traps, sodium vapor traps, a sodium tank. Atomized water-nitrogen process was used for cleaning dummy fuel assembly for CEFR and a sintered stainless steel filter. Sprinkle process was used for cleaning some tubes. Bultylcellosolve was used for cleaning sintered stainless steel filter and sodium flow measurement device. Ethanol alcohol was used for cleaning electromagnetic pump. Paraffin, transformer-oil or their mixture was used for cleaning sodium valves, a sodium vapor trap and sodium-potassium alloy absorber. A small sintered stainless steel filter was distillated in vacuum. A simple sodium disposal device has been served for several years in CIA.E. It can dispose about 10 Kg sodium each time and the disposal process is no-aerosol. It operates in open air for non-radioactive sodium. In recent years a small sodium cleaning plant has been built. It can use atomized water, steam or organic alcohol to removal of sodium. The LAVEL cleaning plant and SLAPSO cleaning plant were introduced from Italy. And CEFR preliminary design on sodium cleaning for spent fuel assembly and on sodium removal-decontamination for large reactor components is introduced. Vapour-nitrogen process is planned to use in them. (author)

  18. Heat transfer system

    Science.gov (United States)

    Not Available

    1980-03-07

    A heat transfer system for a nuclear reactor is described. Heat transfer is accomplished within a sealed vapor chamber which is substantially evacuated prior to use. A heat transfer medium, which is liquid at the design operating temperatures, transfers heat from tubes interposed in the reactor primary loop to spaced tubes connected to a steam line for power generation purposes. Heat transfer is accomplished by a two-phase liquid-vapor-liquid process as used in heat pipes. Condensible gases are removed from the vapor chamber through a vertical extension in open communication with the chamber interior.

  19. Technetium removal from aqueous wastes

    International Nuclear Information System (INIS)

    Fletcher, P.A.; Jones, C.P.; Junkison, A.R.; Turner, A.D.; Kavanagh, P.R.

    1992-03-01

    The research discussed in this report has compared several ''state of the art'' techniques for the removal of traces of the radionuclide, technetium, from aqueous wastes. The techniques investigated were: electrochemical reduction to an insoluble oxide, electrochemical ion exchange, seeded ultrafiltration and chemical reduction followed by filtration. Each technique was examined using a simulant based upon the waste generated by the Enhanced Actinide Removal Plant (EARP) at Sellafield. The technique selected for further investigation was direct electrochemical reduction which offers an ideal route for the removal of technetium from the stream (DFs 10-100) and can be operated continuously with a low power consumption 25 kW for the waste generated by EARP. Cell designs for scale up have been suggested to treat the 1000m 3 of waste produced every day. Future work is proposed to investigate the simultaneous removal of other key radionuclides, such as ruthenium, plutonium and cobalt as well as scale up of the resulting process and to investigate the effect of these other radionuclides on the efficiency of the electrochemical reduction technique for the removal of technetium. Total development and full scale plant costs are estimated to be of the order of 5 pounds - 10M, with a time scale of 5 -8 years to realisation. (author)

  20. Synthesis of palm oil empty fruit bunch magnetic pyrolytic char impregnating with FeCl3 by microwave heating technique

    International Nuclear Information System (INIS)

    Mubarak, N.M.; Kundu, A.; Sahu, J.N.; Abdullah, E.C.; Jayakumar, N.S.

    2014-01-01

    Empty fruit bunch (EFB) is one of the most abundant residues of the Palm oil mill industry in Malaysia. The novel magnetic bio-char was synthesized by single stage microwave heating technique, using EFB in the presence of ferric chloride hexahydrate. The effect of microwave powers, radiation time and impregnation ratio (IR) of ferric chloride hexahydrate to biomass were studied. Also the process parameters such as microwave powers, radiation times and IR were optimized using response surface method. The statistical analysis revealed that the optimum conditions for the high porosity magnetic bio-char production were at 900 W microwave power, 20 min radiation time and 0.5 (FeCl 3 : biomass) impregnation ratio. These newly produced magnetic bio-char have a high surface area of 890 m 2  g −1 and that leads to highly efficient in the removal of methylene blue (MB) with an efficiency of 99.9% from aqueous solution with a maximum adsorption capacity of 265 mg g −1 . - Highlights: • Magnetic bio-char production using discarded material EFB with chemical activation. • Single stage synthesis of magnetic bioc-har via microwave heating was narrated. • Effect of each process parameters on synthesis of magnetic bio-char was elaborated. • Magnetic bio-char has high surface area, high porosity and high adsorption capacity. • Novel magnetic bio-char adds new dimension to the materials as an adsorbent

  1. A standalone decay heat removal device for the Gas-cooled Fast Reactor for intermediate to atmospheric pressure conditions

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A., E-mail: aaron@epiney.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland); Alpy, N., E-mail: nicolas.alpy@cea.fr [CEA, DEN, Service d' Etudes des Systemes Innovants, F-13108 Saint Paul Lez Durance (France); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Chawla, R., E-mail: rakesh.chawla@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer An analytical model predicting Brayton cycle off-design steady states, is developed. Black-Right-Pointing-Pointer The model is used to design an autonomous decay heat removal system for the GFR. Black-Right-Pointing-Pointer Predictions of the analytical model are verified using CATHARE. Black-Right-Pointing-Pointer CATHARE code is used to simulate a set of GFR safety depressurization transients using this device. Black-Right-Pointing-Pointer Convenient turbo-machine designs exist for the targeted autonomous decay heat removal for a wide pressure range. - Abstract: This paper reports a design study for a Brayton cycle machine, which would constitute a dedicated, standalone decay heat removal (DHR) device for the Generation IV Gas-cooled Fast Reactor (GFR). In comparison to the DHR reference strategy developed by the French Commissariat a l'Energie Atomique during the GFR pre-conceptual design phase (which was completed at the end of 2007), the salient feature of this alternative device would be to combine the energetic autonomy of the natural convection process - which is foreseen for operation at high and medium pressures - with the efficiency of the forced convection process which is foreseen for operation down to very low pressures. An analytical model, the so-called 'Brayton scoping model', is described first. This is based on simplified thermodynamic and aerodynamic equations, and was developed to highlight design choices. Two different machine designs are analyzed: a Brayton loop turbo-machine working with helium, and a second one working with nitrogen, since nitrogen is the heavy gas foreseen to be injected into the primary system to enhance the natural convection under loss-of-coolant-accident (LOCA) conditions. Simulations of the steady-state and transient behavior of the proposed device have then been carried out using the CATHARE code. These serve to confirm the insights obtained from usage of the

  2. Nail-gun injury of the cervical spine: simple technique for removal of a barbed nail.

    Science.gov (United States)

    Nathoo, Narendra; Sarkar, Atom; Varma, Gandhi; Mendel, Ehud

    2011-07-01

    Although nail-gun injuries are a common form of penetrating low-velocity injury, impalement with barbed nails has been underreported to date. Barbed nails are designed to resist dislodgment once embedded, and any attempt at removal may splay open the barbs along the path of entry, with the potential for significant soft-tissue and neurovascular injury. A 25-year-old man sustained a nail impalement of the cervical spine from accidental discharge of a nail gun. The patient was noted to be fully conscious with no neurological deficits. Cervical Zone 2 impalement was noted, with only the head of the nail visible. Angiography revealed the nail lying just anterior to the right vertebral artery (VA), with compression of the vessel. Preoperatively, analysis of a similar nail revealed that orientation of the head determined position of the barbs. A deep neck dissection was then performed to the lateral aspect of the C-3 body, using the nail as a guide. Prior to removal, the nail was turned 180° to change the position of the barbs, to prevent injury to the VA. Nail removal was uneventful. The authors present a simple technique for treatment of a nail-gun injury with a barbed nail. Prior to removal, radiographic analysis of the impaled nail must be performed to determine the presence of barbs. If possible, the surgeon should request a similar nail for analysis prior to surgery. Last, the treating surgeon must have knowledge of the barbs' position at all times during nail removal, to prevent damage to critical structures.

  3. FY 1986 Report on research and development of super heat pump energy accumulation system. Part 2. Development of elementary techniques; 1986 nendo super heat pump energy shuseki system no kenkyu kaihatsu seika hokokusho. 2. Yoso gijutsu no kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-04-01

    Summarized in detail herein are R and D results of the chemical heat storage techniques and plant simulation, for R and D of the super heat pump energy accumulation system. For R and D of the chemical heat storage techniques, the R and D efforts are directed to the researches on the fundamental reactions and continuous exothermic reactions involved for the high temperature heat storage type (utilizing the metathesis reactions); researches on the physical properties, heat storage systems, solid-phase reactions, liquid-phase reactors, corrosion of the materials, and so on for the high temperature heat storage type (utilizing ammonia complex); collection of the data related to media and structural materials, tests of the elementary equipment for the absorption and hydration reactions, and so on for the high temperature heat storage type (chemical heat storage utilizing hydration); researches on the media properties and system performance, tests of equipment, and so on for the high temperature heat storage type (heat storage/heating utilizing solvation); researches on the heat storage media, heat storage techniques, corrosion of the materials, systems, and so on for the low temperature heat storage type (utilizing the hydration reactions by mixing solutes); and researches on the media, corrosion and elementary equipment, optimization of the system, and so on for the low temperature heat storage type (clathrate low temperature heat storage systems). (NEDO)

  4. Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator

    Energy Technology Data Exchange (ETDEWEB)

    Burgazzi, Luciano, E-mail: luciano.burgazzi@enea.it

    2014-12-15

    The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing CCFs can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

  5. Six-phase soil heating accelerates VOC extraction from clay soil

    International Nuclear Information System (INIS)

    Gauglitz, P.A.; Roberts, J.S.; Bergsman, T.M.; Caley, S.M.; Heath, W.O.; Miller, M.C.; Moss, R.W.; Schalla, R.; Jarosch, T.R.; Eddy-Dilek, C.A.

    1994-08-01

    Six-Phase Soil Heating (SPSH) was demonstrated as a viable technology for heating low permeability soils containing volatile organic contaminants. Testing was performed as part of the Volatile Organic Compounds in Non-Arid Soils Integrated Demonstration (VOC Non-Arid ID) at the Savannah River Site. The soil at the integrated demonstration site is contaminated with perchloroethylene (PCE) and trichloroethylene (TCE); the highest soil contamination occurs in clay-rich zones that are ineffectively treated by conventional soil vapor extraction due to the very low permeability of the clay. The SPSH demonstration sought to heat the clay zone and enhance the performance of conventional soil vapor extraction. Thermocouples at thirty locations quantified the areal and vertical heating within the treated zone. Soil samples were collected before and after heating to quantify the efficacy of heat-enhanced vapor extraction of PCE and TCE from the clay soil. Samples were taken (essentially every foot) from six wells prior to heating and adjacent to these wells after heating. Results show that contaminant removal from the clay zone was 99.7% (median) within the electrode array. Outside the array where the soil was heated, but to only 50 degrees C, the removal efficiency was 93%, showing that heating accelerated the removal of VOCs from the clay soil. The accelerated remediation resulted from effective heating of the contaminated clay zone by SPSH. The temperature of the clay zone increased to 100 degrees C after 8 days of heating and was maintained near 100 degrees C for 17 days. Electrical heating removed 19,000 gal of water from the soil as steam, with peak removal rate of 1,500 gpd of condensed steam

  6. Removal of Pb(II) from water by the activated carbon modified by nitric acid under microwave heating.

    Science.gov (United States)

    Yao, Shuheng; Zhang, Jiajun; Shen, Dekui; Xiao, Rui; Gu, Sai; Zhao, Ming; Liang, Junyu

    2016-02-01

    The rice husk based activated carbon (RH-AC) was treated by nitric acid under microwave heating, in order to improve its capability for the removal of heavy metal ions from water. The optimal conditions for the modification of RH-AC (M-RH-AC) were determined by means of orthogonal array experimental design, giving those as the concentration of nitric acid of 8mol/L, modification time of 15min, modification temperature of 130°C and microwave power of 800W. The characteristics of the M-RH-AC and RH-AC were examined by BET, XRD, Raman spectrum, pH titration, zeta potential, Boehm titration and FTIR analysis. The M-RH-AC has lower pore surface area, smaller crystallite, lower pHIEP and more oxygen-containing functional groups than the RH-AC. Removal capacity of Pb(II) ions by the M-RH-AC and RH-AC from water solution was estimated concerning the influence of contact time, pH value, and initial concentration. The equilibrium time of Pb(II) removal was found to be around 90min after modification process. Two kinetic models are adopted to describe the possible Pb(II) adsorption mechanism, finding that the adsorption rate of Pb(II) ions by the M-RH-AC is larger than that of RH-AC. Copyright © 2015 Elsevier Inc. All rights reserved.

  7. Test Plan for the overburden removal demonstration

    International Nuclear Information System (INIS)

    Rice, P.; Thompson, D.; Winberg, M.; Skaggs, J.

    1993-06-01

    The removal of soil overburdens from contaminated pits and trenches involves using equipment that will remove a small layer of soil from 3 to 6 in. at any time. As a layer of soil is removed, overburden characterization techniques perform surveys to a depth that exceeds each overburden removal layer to ensure that the removed soil will be free of contamination. It is generally expected that no contamination will be found in the soil overburden, which was brought in after the waste was put in place. It is anticipated that some containers in the waste zone have lost their integrity, and the waste leakage from those containers has migrated by gravity downward into the waste zone. To maintain a safe work environment, this method of overburden removal should allow safe preparation of a pit or trench for final remediation. To demonstrate the soil overburden techniques, the Buried Waste Integrated Demonstration Program has contracted vendor services to provide equipment and techniques demonstrating soil overburden removal technology. The demonstration will include tests that will evaluate equipment performance and techniques for removal of overburden soil, control of contamination spread, and dust control. To evaluate the performance of these techniques, air particulate samples, physical measurements of the excavation soil cuts, maneuverability measurements, and time versus volume (rate) of soil removal data will be collected during removal operations. To provide a medium for sample evaluation, the overburden will be spiked at specific locations and depths with rare earth tracers. This test plan will be describe the objectives of the demonstration, data quality objectives, methods to be used to operate the equipment and use the techniques in the test area, and methods to be used in collecting data during the demonstration

  8. Experimental study of natural two-phase flow circulation using a visualization technique

    International Nuclear Information System (INIS)

    Vinhas, Pedro A.M.; Su, Jian

    2013-01-01

    This paper presents an experimental study of natural two-phase flow in a circuit that simulates, on a smaller scale, a typical residual heat removal system of passive reactors APWR (Advanced Pressurized Water Reactor). The circuit was formed by a heater, a heat exchanger and piping. The experimental study was the application of a visualization technique, using a high speed camera, for measuring the size and speed of vapor bubbles generated in the heater with different power heating. The camera was positioned in the central region of the pipe connecting the heater to the heat exchanger, where there is a clear passage. The flow of images were processed and analyzed using commercial software that allowed the determination of the length and velocity of the bubbles. The results were then compared with correlations available in literature

  9. Graphite Foam Heat Exchangers for Thermal Management

    Energy Technology Data Exchange (ETDEWEB)

    Klett, J.W.

    2004-06-07

    -bond{reg_sign}, but still better than the standard heat sinks. Next, work with evaporative cooling techniques, such as heat pipes, demonstrated some unique behavior with the foam that is not seen with standard wick materials. This was that as the thickness of the foam increased, the performance got better, where with standard wick materials, as the thickness increases, the performance decreases. This is yet to be completely explained. Last, the designs from the thermal model were used to fabricate a series of cold plates with the graphite foam and compare them to similar designs using high performance folded fin aluminum sinks (considered standard in the industry). It was shown that by corrugating the foam parallel to fluid flow, the pressure drop can be reduced significantly while maintaining the same heat transfer as that in the folded fin heat sink. In fact, the results show that the graphite foam heat sink can utilized 5% the pumping power as that required with the folded fin aluminum heat sink, yet remove the same amount of heat.

  10. Modelling of Evaporator in Waste Heat Recovery System using Finite Volume Method and Fuzzy Technique

    Directory of Open Access Journals (Sweden)

    Jahedul Islam Chowdhury

    2015-12-01

    Full Text Available The evaporator is an important component in the Organic Rankine Cycle (ORC-based Waste Heat Recovery (WHR system since the effective heat transfer of this device reflects on the efficiency of the system. When the WHR system operates under supercritical conditions, the heat transfer mechanism in the evaporator is unpredictable due to the change of thermo-physical properties of the fluid with temperature. Although the conventional finite volume model can successfully capture those changes in the evaporator of the WHR process, the computation time for this method is high. To reduce the computation time, this paper develops a new fuzzy based evaporator model and compares its performance with the finite volume method. The results show that the fuzzy technique can be applied to predict the output of the supercritical evaporator in the waste heat recovery system and can significantly reduce the required computation time. The proposed model, therefore, has the potential to be used in real time control applications.

  11. Three Incomplete Caries Removal Techniques Compared Over Two Years in Primary Molars with Asymptomatic Deep Caries or Reversible Pulpitis.

    Science.gov (United States)

    Chompu-inwai, Papimon; Boonsongsawat, Kamolthip; Sastraruji, Thanapat; Sophasri, Tidarat; Mankaen, Siripun; Nondon, Sutasinee; Tunlek, Sumattaya; Katwong, Supitchaya

    2015-01-01

    To directly compare the survival rates of three incomplete caries removal techniques that differed in the amount of caries removal and the base material used. Ninety-six primary molars with asymptomatic deep caries or reversible pulpitis were randomly assigned to three groups: (1) indirect pulp treatment (IPT); (2) minimal caries removal with both resin-modified glass ionomer base material and luting cement (MCRB/L); and (3) minimal caries removal with only resin-modified glass ionomer luting cement (MCRL). The treatments were followed clinically and radiographically for two years. The two-year survival probabilities in the IPT, MCRB/L, and MCRL groups were 0.90 (95 percent confidence interval [CI] equals 0.73 to 0.97), 0.93 (95 percent CI equals 0.76 to 0.98), and 0.77 (95 percent CI equals 0.58 to 0.89), respectively. There was no significant difference in the two-year survival probabilities of the three studied groups (generalized Wilcoxon P=.07). Following two years, neither the amount of caries removal nor the base material affected the success of incomplete caries removal treatment. However, minimal caries removal with MCRB/L presented the highest survival rate among the tested groups and resulted in no incidence of pulp exposure.

  12. A simple mathematical procedure to estimate heat flux in machining using measured surface temperature with infrared laser

    Directory of Open Access Journals (Sweden)

    Hocine Mzad

    2015-09-01

    Full Text Available Several techniques have been developed over time for the measurement of heat and the temperatures generated in various manufacturing processes and tribological applications. Each technique has its own advantages and disadvantages. The appropriate technique for temperature measurement depends on the application under consideration as well as the available tools for measurement. This paper presents a procedure for a simple and accurate determination of the time-varying heat flux at the workpiece–tool interface of three different metals under known cutting conditions. A portable infrared thermometer is used for surface temperature measurements. A spline smoothing interpolation of the surface temperature history enables to determine the local heat flux produced during stock removal. The measured temperature is represented by a third-order spline approximation. Nonetheless, the accuracy of polynomial interpolation depends on how close are the interpolated points; an increase in degree cannot be used to increase the accuracy. Although the data analysis is relatively complicated, the computing time is very small.

  13. Fluoroscopic removal of retrievable self-expandable metal stents in patients with malignant oesophageal strictures: Experience with a non-endoscopic removal system.

    Science.gov (United States)

    Kim, Pyeong Hwa; Song, Ho-Young; Park, Jung-Hoon; Zhou, Wei-Zhong; Na, Han Kyu; Cho, Young Chul; Jun, Eun Jung; Kim, Jun Ki; Kim, Guk Bae

    2017-03-01

    To evaluate clinical outcomes of fluoroscopic removal of retrievable self-expandable metal stents (SEMSs) for malignant oesophageal strictures, to compare clinical outcomes of three different removal techniques, and to identify predictive factors of successful removal by the standard technique (primary technical success). A total of 137 stents were removed from 128 patients with malignant oesophageal strictures. Primary overall technical success and removal-related complications were evaluated. Logistic regression models were constructed to identify predictive factors of primary technical success. Primary technical success rate was 78.8 % (108/137). Complications occurred in six (4.4 %) cases. Stent location in the upper oesophagus (P=0.004), stricture length over 8 cm (P=0.030), and proximal granulation tissue (Pstent location in the upper oesophagus, and stricture length over 8 cm were negative predictive factors for primary technical success by standard extraction and may require a modified removal technique. • Fluoroscopic retrievable SEMS removal is safe and effective. • Standard removal technique by traction is effective in the majority of patients. • Three negative predictive factors of primary technical success were identified. • Caution should be exercised during the removal in those situations. • Eversion technique is effective in cases of proximal granulation tissue.

  14. Experimental investigations on scaled models for the SNR-2 decay heat removal by natural convection

    International Nuclear Information System (INIS)

    Hoffmann, H.; Weinberg, D.; Tschoeke, H.; Frey, H.H.; Pertmer, G.

    1986-01-01

    Scaled water models are used to prove the mode of function of the decay heat removal by natural convection for the SNR-2. The 2D and 3D models were designed to reach the characteristic numbers (Richardson, Peclet) of the reactor. In the experiments on 2D models the position of the immersed cooler (IC) and the power were varied. Temperature fields and velocities were measured. The IC installed as a separate component in the hot plenum resulted in a very complex flow behavior and low temperatures. Integrating the IC in the IHX showed a very simple circulating flow and high temperatures within the hot plenum. With increasing power only slightly rising temperature differences within the core and IC were detected. Recalculations using the COMMIX 1B code gave qualitatively satisfying results. (author)

  15. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    International Nuclear Information System (INIS)

    1980-01-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases

  16. Evaluation of the Safety Issue Concerning the Potential for Loss of Decay Heat Removal Function due to Crude Oil Spill in the Ultimate Heat Sink of Nuclear Reactors

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Roh, Kyung Wan; Yune, Young Gill; Kang, Dong Gu; Kim, Hho Jhung

    2008-01-01

    A barge crashed into a moored oil tanker at about 7:15 a.m., Dec. 12, 2007, dumping around 10,500 tons of crude oil into the sea in Korea. The incident took place about 15 kilometers northwest of Manripo beach in South Chungcheong where is Korea's west coast in the Yellow Sea. In a few days, the oil slicks spread to the northern and southern tips of the Taean Peninsula by strong winds and tides. As time went the spilled oil floating on the surface of sea water was volatilized to become tar-balls and lumps and drifted far away in the southern direction. 13 days after the incident, some of oil slicks and tar lumps were observed to flow in the service water intake at the Younggwang nuclear power plants (NPPs) operating 6 reactors, which are over 150 km away from the incident spot in the southeastern direction. According to the report by the Younggwang NPPs, a total weight 83 kg of tar lumps was removed for about 3 days. Oil spills in the sea can happen in any country or anytime due to human errors or mistakes, wars, terrors, intentional dumping of waste oils, and natural disasters like typhoon and tsunami. In fact, there have been 7 major oil spills over 10,000 tons that have occurred around the world since 1983. As such serious oil spill incidents may happen near the operating power plants using the sea water as ultimate heat sink. To ensure the safe operation of nuclear reactors it is required to evaluate the potential for loss of decay heat removal function of nuclear reactors due to the spilled oils flowing in the service water intake, from which the service water is pumped. Thus, Korea Institute of Nuclear Safety identified this problem as one of the important safety. When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where all service water pumps locate, by sea current and wind drift (induced) current. The essential service water pumps take the service

  17. Evaluation of the Safety Issue Concerning the Potential for Loss of Decay Heat Removal Function due to Crude Oil Spill in the Ultimate Heat Sink of Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Roh, Kyung Wan; Yune, Young Gill; Kang, Dong Gu; Kim, Hho Jhung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    A barge crashed into a moored oil tanker at about 7:15 a.m., Dec. 12, 2007, dumping around 10,500 tons of crude oil into the sea in Korea. The incident took place about 15 kilometers northwest of Manripo beach in South Chungcheong where is Korea's west coast in the Yellow Sea. In a few days, the oil slicks spread to the northern and southern tips of the Taean Peninsula by strong winds and tides. As time went the spilled oil floating on the surface of sea water was volatilized to become tar-balls and lumps and drifted far away in the southern direction. 13 days after the incident, some of oil slicks and tar lumps were observed to flow in the service water intake at the Younggwang nuclear power plants (NPPs) operating 6 reactors, which are over 150 km away from the incident spot in the southeastern direction. According to the report by the Younggwang NPPs, a total weight 83 kg of tar lumps was removed for about 3 days. Oil spills in the sea can happen in any country or anytime due to human errors or mistakes, wars, terrors, intentional dumping of waste oils, and natural disasters like typhoon and tsunami. In fact, there have been 7 major oil spills over 10,000 tons that have occurred around the world since 1983. As such serious oil spill incidents may happen near the operating power plants using the sea water as ultimate heat sink. To ensure the safe operation of nuclear reactors it is required to evaluate the potential for loss of decay heat removal function of nuclear reactors due to the spilled oils flowing in the service water intake, from which the service water is pumped. Thus, Korea Institute of Nuclear Safety identified this problem as one of the important safety. When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where all service water pumps locate, by sea current and wind drift (induced) current. The essential service water pumps take the

  18. Targeting the maximum heat recovery for systems with heat losses and heat gains

    International Nuclear Information System (INIS)

    Wan Alwi, Sharifah Rafidah; Lee, Carmen Kar Mun; Lee, Kim Yau; Abd Manan, Zainuddin; Fraser, Duncan M.

    2014-01-01

    Graphical abstract: Illustration of heat gains and losses from process streams. - Highlights: • Maximising energy savings through heat losses or gains. • Identifying location where insulation can be avoided. • Heuristics to maximise heat losses or gains. • Targeting heat losses or gains using the extended STEP technique and HEAT diagram. - Abstract: Process Integration using the Pinch Analysis technique has been widely used as a tool for the optimal design of heat exchanger networks (HENs). The Composite Curves and the Stream Temperature versus Enthalpy Plot (STEP) are among the graphical tools used to target the maximum heat recovery for a HEN. However, these tools assume that heat losses and heat gains are negligible. This work presents an approach that considers heat losses and heat gains during the establishment of the minimum utility targets. The STEP method, which is plotted based on the individual, as opposed to the composite streams, has been extended to consider the effect of heat losses and heat gains during stream matching. Several rules to guide the proper location of pipe insulation, and the appropriate procedure for stream shifting have been introduced in order to minimise the heat losses and maximise the heat gains. Application of the method on two case studies shows that considering heat losses and heat gains yield more realistic utility targets and help reduce both the insulation capital cost and utility cost of a HEN

  19. Transient Analysis of a Magnetic Heat Pump

    Science.gov (United States)

    Schroeder, E. A.

    1985-01-01

    An experimental heat pump that uses a rare earth element as the refrigerant is modeled using NASTRAN. The refrigerant is a ferromagnetic metal whose temperature rises when a magnetic field is applied and falls when the magnetic field is removed. The heat pump is used as a refrigerator to remove heat from a reservoir and discharge it through a heat exchanger. In the NASTRAN model the components modeled are represented by one-dimensional ROD elements. Heat flow in the solids and fluid are analyzed. The problem is mildly nonlinear since the heat capacity of the refrigerant is temperature-dependent. One simulation run consists of a series of transient analyses, each representing one stroke of the heat pump. An auxiliary program was written that uses the results of one NASTRAN analysis to generate data for the next NASTRAN analysis.

  20. Method of extracting heat from dry geothermal reservoirs

    Science.gov (United States)

    Potter, R.M.; Robinson, E.S.; Smith, M.C.

    1974-01-22

    Hydraulic fracturing is used to interconnect two or more holes that penetrate a previously dry geothermal reservoir, and to produce within the reservoir a sufficiently large heat-transfer surface so that heat can be extracted from the reservoir at a usefully high rate by a fluid entering it through one hole and leaving it through another. Introduction of a fluid into the reservoir to remove heat from it and establishment of natural (unpumped) convective circulation through the reservoir to accomplish continuous heat removal are important and novel features of the method. (auth)

  1. Analysis of removal of residual decay heat from interim storage facilities by means of the CFD program FLUENT

    International Nuclear Information System (INIS)

    Stratmann, W.; Hages, P.

    2004-01-01

    Within the scope of nuclear licensing procedures of on-site interim storage facilities for dual purpose casks it is necessary, among other things, to provide proof of sufficient removal of the residual decay heat emitted by the casks. The results of the analyses performed for this purpose define e.g. the boundary conditions for further thermal analyses regarding the permissible cask component temperatures or the maximum permissible temperatures of the fuel cladding tubes of the fuel elements stored in the casks. Up to now, for the centralized interim storage facilities in Germany such analyses were performed on the basis of experimental investigations using scaled-down storage geometries. In the engineering phase of the Lingen on-site interim storage facility, proof was furnished for the first time using the CFD (computational fluid dynamics) program FLUENT. The program FLUENT is an internationally recognized and comprehensively verified program for the calculation of flow and heat transport processes. Starting from a brief discussion of modeling and the different boundary conditions of the computation, this contribution presents various results regarding the temperatures of air, cask surfaces and storage facility components, the mass flows through the storage facility and the heat transfer at the cask surface. The interface point to the cask-specific analyses is defined to be the cask surface

  2. Experimental evaluation of sodium to air heat exchanger performance

    International Nuclear Information System (INIS)

    Vinod, V.; Pathak, S.P.; Paunikar, V.D.; Suresh Kumar, V.A.; Noushad, I.B.; Rajan, K.K.

    2013-01-01

    Highlights: ► Sodium to air heat exchangers are used to remove the decay heat produced in fast breeder reactor after shutdown. ► Finned tube sodium to air heat exchanger with sodium on tube side was tested for its heat transfer performance. ► A one dimensional computer code was validated by the experimental data obtained. ► Non uniform sodium and air flow distribution was present in the heat exchanger. - Abstract: Sodium to air heat exchangers (AHXs) is used in Prototype Fast Breeder Reactor (PFBR) circuits to reject the decay heat produced by the radioactive decay of the fission products after reactor shutdown, to the atmospheric air. The heat removal through sodium to air heat exchanger maintains the temperature of reactor components in the pool within safe limits in case of non availability of normal heat transport path. The performance of sodium to air heat exchanger is very critical to ensure high reliability of the decay heat removal systems in sodium cooled fast breeder reactors. Hence experimental evaluation of the adequacy of the heat transfer capability gives confidence to the designers. A finned tube cross flow sodium to air heat exchanger of 2 MW heat transfer capacity with sodium on tube side and air on shell side was tested in the Steam Generator Test Facility at Indira Gandhi Center for Atomic Research, India. Heat transfer experiments were carried out with forced circulation of sodium and air, which confirmed the adequacy of heat removal capacity of the heat exchanger. The testing showed that 2.34 MW of heat power is transferred from sodium to air at nominal flow and temperature conditions. A one dimensional computer code developed for design and analysis of the sodium to air heat exchanger was validated by the experimental data obtained. An equivalent Nusselt number, Nu eq is derived by approximating that the resistance of heat transfer from sodium to air is contributed only by the film resistance of air. The variation of Nu eq with respect

  3. Graphene-enhanced thermal interface materials for heat removal from photovoltaic solar cells

    Science.gov (United States)

    Saadah, M.; Gamalath, D.; Hernandez, E.; Balandin, A. A.

    2016-09-01

    The increase in the temperature of photovoltaic (PV) solar cells affects negatively their power conversion efficiency and decreases their lifetime. The negative effects are particularly pronounced in concentrator solar cells. Therefore, it is crucial to limit the PV cell temperature by effectively removing the excess heat. Conventional thermal phase change materials (PCMs) and thermal interface materials (TIMs) do not possess the thermal conductivity values sufficient for thermal management of the next generation of PV cells. In this paper, we report the results of investigation of the increased efficiency of PV cells with the use of graphene-enhanced TIMs. Graphene reveals the highest values of the intrinsic thermal conductivity. It was also shown that the thermal conductivity of composites can be increased via utilization of graphene fillers. We prepared TIMs with up to 6% of graphene designed specifically for PV cell application. The solar cells were tested using the solar simulation module. It was found that the drop in the output voltage of the solar panel under two-sun concentrated illumination can be reduced from 19% to 6% when grapheneenhanced TIMs are used. The proposed method can recover up to 75% of the power loss in solar cells.

  4. Heat exchanger restart evaluation

    International Nuclear Information System (INIS)

    Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

    1992-01-01

    On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summarized

  5. Heat exchanger restart evaluation

    International Nuclear Information System (INIS)

    Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

    1992-01-01

    On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summary herein

  6. Neutron studies of nanostructured CuO-Al2O3 NOx removal catalysts

    International Nuclear Information System (INIS)

    Ozawa, Masakuni; Loong Chun-Keung

    1997-01-01

    Nanostructured powders of automotive catalytic system CuO0Al 2 O 3 , targeted for nitrogen oxides (NOx) removal under lean-burn engine conditions, were investigated using neutron diffraction and small-angle neutron scattering. The crystal phases, structural transformations and microstructure of 10 mol% Cu-Al 2 O 3 powders are characterized according to the heat-treatment conditions. These properties are correlated with the pore structure and NOx removal efficiency determined by nitrogen adsorption isotherm, electron spin resonance, and temperature programmed reaction measurements. The γ-(Cu, Al) 2 O 3 phase and the mass-fractal-like aggregate of particles (size ∼ 26 nm) at annealing temperatures below 900 degrees C were found to be crucial to the high NOx removal performance. The transformation to bulk crystalline phases of α-Al 2 O 3 + CuAl 2 O 4 spinel above ∼1050 degrees C corresponds to a drastic drop of Nox removal efficiency. The usefulness of neutron-scattering techniques as well as their complementarity with other traditional methods of catalytic research are discussed

  7. New Technique for Cryogenically Cooling Small Test Articles

    Science.gov (United States)

    Rodriquez, Karen M.; Henderson, Donald J.

    2011-01-01

    Convective heat removal techniques to rapidly cool small test articles to Earth-Moon L2 temperatures of 77 K were accomplished through the use of liquid nitrogen (LN2). By maintaining a selected pressure range on the saturation curve, test articles were cooled below the LN2 boiling point at ambient pressure in less than 30 min. Difficulties in achieving test pressures while maintaining the temperature tolerance necessitated a modification to the original system to include a closed loop conductive cold plate and cryogenic shroud

  8. Critical heat fluxes in tubular fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    Subbotin, V.I.; Alekseev, G.V.; Peskov, O.L.

    1974-01-01

    The results of the experiments carried out show that with appropriate choice of tube, type and dimensions of intensifier the attainment of critical conditions at certain parameters is not accompanied by sharp or considerable increases in temperature of the heat removing surface. Increase in power to above critical under these conditions does not lead to considerable variation in temperature either. Thus, it appears possible to change from heat removal by steam-water mixture to convective heat removal by wet steam without manifestation of intolerable temperature conditions of the heating surface (Fig. 6). A change to convective heat removal by wet steam is possible at different levels of heat fluxes which depend during constant conditions at the inlet on tube length and the degree of the disturbing influence on the flow. This is especially important since in principle the possibility arises for developing a power reactor with tubular fuel elements, in which a once-through cycle with steam superheat involving no intermediate separation can be realised

  9. Dislodgement and removal of dust-particles from a surface by a technique combining acoustic standing wave and airflow.

    Science.gov (United States)

    Chen, Di; Wu, Junru

    2010-01-01

    It is known that there are many fine particles on the moon and Mars. Their existence may cause risk for the success of a long-term project for NASA, i.e., exploration and habitation of the moon and Mars. These dust-particles might cover the solar panels, making them fail to generate electricity, and they might also penetrate through seals on space suits, hatches, and vehicle wheels causing many incidents. The fine particles would be hazardous to human health if they were inhaled. Development of robust dust mitigation technology is urgently needed for the viable long-term exploration and habilitation of either the moon or Mars. A feasibility study to develop a dust removal technique, which may be used in space-stations or other enclosures for habitation, is reported. It is shown experimentally that the acoustic radiation force produced by a 13.8 kHz 128 dB sound-level standing wave between a 3 cm-aperture tweeter and a reflector separated by 9 cm is strong enough to overcome the van der Waals adhesive force between the dust-particles and the reflector-surface. Thus the majority of fine particles (>2 microm diameter) on a reflector-surface can be dislodged and removed by a technique combining acoustic levitation and airflow methods. The removal efficiency deteriorates for particles of less than 2 microm in size.

  10. Cyro Power and Heat Transfer

    National Research Council Canada - National Science Library

    Chow, L

    1998-01-01

    .... The heat generated from a 9x9-heater array was removed by liquid nitrogen pool boiling. The orientation and space limitation of the array were varied to explore their effects on the critical heat flux (CHF) value...

  11. KINETICS PROCESSES OF DEHYDRATION AND HEATING FISH DURING FRYING, DURING SEMI HOT AND HOT SMOKING

    Directory of Open Access Journals (Sweden)

    V. A. Pokholchenko

    2014-01-01

    Full Text Available Summary. Calculated methods of graphing of curves for kinetics of dehydration and fish heating during the processes of frying, semi hot smoking and hot smoking have been developed. The offered methods of calculating are based on the basic regularities of heat and mass exchanges of these processes. Based on the research of the regularities of dehydration on the kinetic curves, critical points were identified, that characterize the transition from the moisture removal with lower energy of its bond with material to the removal of one with higher energy bond, also the influence of the product shrinkage on the velocity of the moisture removal. These points are characteristic for the temperature curves as well. It’s suggested for the temperature curve to be replaced by broken line that consists of three straight lines that are crossing in points, corresponded with the critical moistures and critical temperatures. Significant amount of the experimental material of the research of the kinetics of dehydration and fish heating under different modes is shown by authors in the form of generalized dependencies. The method allows modeling the processes of heating and dehydrating of fish and choosing the most rational modes based on the calculated data. The proposed technique makes it possible to construct the curves of the kinetics of heating and dehydration kinetics in processes of roasting, semi hot and hot smoked fish, which allows to optimize a particular process, design more efficient in terms of consumption of raw materials and energy technology, as well as to create better machines or upgrade existing equipment into account the relationship of heat and mass transfer processes.

  12. Validation of the TASS/SMR-S Code for the PRHRS Condensation Heat Transfer Model

    International Nuclear Information System (INIS)

    Jun, In Sub; Yang, Soo Hyoung; Chung, Young Jong; Lee, Won Jae

    2011-01-01

    When some accidents or events are occurred in the SMART, the secondary system is used to remove the core decay heat for the long time such as a feedwater system. But if the feedwater system can't remove the residual core heat because of its malfunction, the core decay heat is removed using the Passive Residual Heat Removal System (PRHRS). The PRHRS is passive type safety system adopted to enhance the safety of the SMART. It can fundamentally eliminate the uncertainty of operator action. TASS/SMR-S (Transient And Setpoint Simulation/ System-integrated Modular Reactor-Safety) code has various heat transfer models reflecting the design features of the SMART. One of the heat transfer models is the PRHRS condensation heat transfer model. The role of this model is to calculate the heat transfer coefficient in the heat exchanger (H/X) tube side using the relevant heat transfer correlations for all of the heat transfer modes. In this paper, the validation of the condensation heat transfer model was carried out using the POSTECH H/X heat transfer test

  13. Electrochemical filtration for turbidity removal in industrial cooling/process water systems

    International Nuclear Information System (INIS)

    Kumbhar, A.G.; Venkateswaran, G.

    2008-01-01

    Water samples of large cooling water reservoirs may look visibly clear and transparent, but still may contain sub-micron size particles at sub-parts-per-million levels. Deposition of these particles on heat exchanger surfaces, reduces the heat transfer efficiency in power industry. In nuclear power plants, additionally it creates radiation exposure problems due to activation of fine metallic turbidity in the reactor core and its subsequent transfer to out-of-core surfaces. Sub-micron filtration creates back high-pressure problem. Zeta filters available commercially are prescribed for separating either positively or negatively charged particles. They are of once-use and throw-type. Precipitation surface modified ion exchangers impart chemical impurities to the system. Thus, sub-micron size and dilute turbidity removal from large volumes of waters such as heat exchanger cooling water in nuclear and power industry poses a problem. Electro deposition of the turbidity causing particles, on porous carbon/graphite felt electrodes, is one of the best suited methods for turbidity removal from large volumes of water due to the filter's high permeability, inertness to the system and regenerability resulting in low waste generation. Initially, active indium turbidity removal from RAPS-1 heavy water moderator system, and microbes removal from heat exchanger cooling lake water of RAPS 1 and 2 were demonstrated with in-house designed and fabricated prototype electrochemical filter (ECF). Subsequently, a larger size, high flow filter was fabricated and deployed for iron turbidity removal from active process waters system of Kaiga Generation Station unit 1 and silica and iron turbidity removal from cooling water pond used for heat exchanger of a high temperature high pressure (HTHP) loop at WSCD, Kalpakkam. The ECF proved its exclusive utility for sub-micron size turbidity removal and microbes removal. ECF maneuverability with potential and current for both positively and

  14. Analysis of the Integral Response of CAREM Reactor and the Residual Heat Removal System During a Failure of the Steam Generators Feed Water System

    International Nuclear Information System (INIS)

    Gimenez, Marcelo; Zanocco, Pablo; Schlamp, Miguel

    2000-01-01

    A global analysis of the behavior of Carem-25 Reactor and Residual Heat Removal System (RHRS) to mitigate a loss of heat sink accident is done in the present work.The proposed RHRS removes 2 MW of power and is duplicated to fulfill the redundancy criteria.It consists of two condensers with two tubes in a parallel array.Each tube has 2 S CH 160 TP 347 SS and 2 m 2 of area.The RHRS design requierements (for this accidental sequence) are: Short-term: primary circuit pressure must remain below the safety valves opening set point and the condensers must not flood in order to avoid instabilities. Long-term: reach hot-shutdown condition (primary circuit pressure below 2.3 MPa) at least before 48 hrs. Short-term reactor behavior is simulated using RELAP5 with a detail nodalization of the primary circuit and RHRS.Long term performance is simulated with a simple and conservative model, assuming a saturated primary circuit. This condition is expected during RHRS operation

  15. Isolation, screening and molecular identification of novel bacterial strain removing methylene blue from water solutions

    Science.gov (United States)

    Kilany, Mona

    2017-11-01

    The potentially deleterious effects of methylene blue (MB) on human health drove the interest in its removal promptly. Bioremediation is an effective and eco friendly for removing MB. Soil bacteria were isolated and examined for their potential to remove MB. The most potent bacterial candidate was characterized and identified using 16S rRNA sequence technique. The evolutionary history of the isolate was conducted by maximum likelihood method. Some physiochemical parameters were optimized for maximum decolorization. Decolorization mechanism and microbial toxicity study of MB (100 mg/l) and by-products were investigated. Participation of heat killed bacteria in color adsorption have been investigated too. The bacterial isolate was identified as Stenotrophomonas maltophilia strain Kilany_MB 16S ribosomal RNA gene with 99% sequence similarity. The sequence was submitted to NCBI (Accession number = KU533726). Phylogeny depicted the phylogenetic relationships between 16S ribosomal RNA gene, partial sequence (1442 bp), of the isolated strain and other strains related to Stenotrophomonas maltophilia in the GenBank database. The optimal conditions were investigated to be pH 5 at 30 °C, after 24 h using 5 mg/l MB showing optimum decolorization percentage (61.3%). Microbial toxicity study demonstrated relative reduction in the toxicity of MB decolorized products on test bacteria. Mechanism of color removal was proved by both biosorption and biodegradation, where heat-killed and live cells showed 43 and 52% of decolorization, respectively, as a maximum value after 24-h incubation. It was demonstrated that the mechanism of color removal is by adsorption. Therefore, good performance of S maltophilia in MB color removal reinforces the exploitation of these bacteria in environmental clean-up and restoration of the ecosystem.

  16. Institute for High Heat Flux Removal (IHHFR). Phases I, II, and III

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, Ronald D. [Prairie View A& M Univ., TX (United States)

    2014-08-31

    The IHHFR focused on interdisciplinary applications as it relates to high heat flux engineering issues and problems which arise due to engineering systems being miniaturized, optimized, or requiring increased high heat flux performance. The work in the IHHFR focused on water as a coolant and includes: (1) the development, design, and construction of the high heat flux flow loop and facility; (2) test section development, design, and fabrication; and, (3) single-side heat flux experiments to produce 2-D boiling curves and 3-D conjugate heat transfer measurements for single-side heated test sections. This work provides data for comparisons with previously developed and new single-side heated correlations and approaches that address the single-side heated effect on heat transfer. In addition, this work includes the addition of single-side heated circular TS and a monoblock test section with a helical wire insert. Finally, the present work includes: (1) data base expansion for the monoblock with a helical wire insert (only for the latter geometry), (2) prediction and verification using finite element, (3) monoblock model and methodology development analyses, and (4) an alternate model development for a hypervapotron and related conjugate heat transfer controlling parameters.

  17. Fabrication of a superhydrophobic surface with fungus-cleaning properties on brazed aluminum for industrial application in heat exchangers

    Science.gov (United States)

    Lee, Jeong-Won; Hwang, Woonbong

    2018-06-01

    Extensive research has been carried out concerning the application of superhydrophobic coating in heat exchangers, but little is known about the application of this technique to brazed aluminum heat exchangers (BAHEs). In this work, we describe a new superhydrophobic coating method, which is suitable for BAHE use on an industrial scale. We first render the BAHE superhydrophobic by fabricating micro/nanostructures using solution dipping followed by fluorination. After the complete removal of the silicon residue, we verify using surface analysis that the BAHE surface is perfectly superhydrophobic. We also studied the fungus-cleaning properties of the superhydrophobic surface by growing fungus for 4 weeks in a moist environment on BAHE fins with and without superhydrophobic coating. We observed that, whereas the fungus grown on the untreated fins is extremely difficult to remove, the fungus on the fins with the superhydrophobic coating can be removed easily with only a modest amount of water. We also found that the coated BAHE fins exhibit excellent resistance to moisture. The superhydrophobic coating method that we propose is therefore expected to have a major impact in the heating, ventilating and air conditioning industry market.

  18. Coupled electrokinetics-adsorption technique for simultaneous removal of heavy metals and organics from saline-sodic soil.

    Science.gov (United States)

    Lukman, Salihu; Essa, Mohammed Hussain; Mu'azu, Nuhu Dalhat; Bukhari, Alaadin

    2013-01-01

    In situ remediation technologies for contaminated soils are faced with significant technical challenges when the contaminated soil has low permeability. Popular traditional technologies are rendered ineffective due to the difficulty encountered in accessing the contaminants as well as when employed in settings where the soil contains mixed contaminants such as petroleum hydrocarbons, heavy metals, and polar organics. In this study, an integrated in situ remediation technique that couples electrokinetics with adsorption, using locally produced granular activated carbon from date palm pits in the treatment zones that are installed directly to bracket the contaminated soils at bench-scale, is investigated. Natural saline-sodic soil, spiked with contaminant mixture (kerosene, phenol, Cr, Cd, Cu, Zn, Pb, and Hg), was used in this study to investigate the efficiency of contaminant removal. For the 21-day period of continuous electrokinetics-adsorption experimental run, efficiency for the removal of Zn, Pb, Cu, Cd, Cr, Hg, phenol, and kerosene was found to reach 26.8, 55.8, 41.0, 34.4, 75.9, 92.49, 100.0, and 49.8%, respectively. The results obtained suggest that integrating adsorption into electrokinetic technology is a promising solution for removal of contaminant mixture from saline-sodic soils.

  19. Numerical investigation of vessel heating using a copper vapor laser and a pulsed dye laser in treating vascular skin lesions

    Science.gov (United States)

    Pushkareva, A. E.; Ponomarev, I. V.; Isaev, A. A.; Klyuchareva, S. V.

    2018-02-01

    A computer simulation technique was employed to study the selective heating of a tissue vessel using emission from a pulsed copper vapor laser and a pulsed dye laser. The depth and size of vessels that could be selectively and safely removed were determined for the lasers under examination.

  20. Heating and thermal control of brazing technique to break contamination path for potential Mars sample return

    Science.gov (United States)

    Bao, Xiaoqi; Badescu, Mircea; Sherrit, Stewart; Bar-Cohen, Yoseph; Campos, Sergio

    2017-04-01

    The potential return of Mars sample material is of great interest to the planetary science community, as it would enable extensive analysis of samples with highly sensitive laboratory instruments. It is important to make sure such a mission concept would not bring any living microbes, which may possibly exist on Mars, back to Earth's environment. In order to ensure the isolation of Mars microbes from Earth's Atmosphere, a brazing sealing and sterilizing technique was proposed to break the Mars-to-Earth contamination path. Effectively, heating the brazing zone in high vacuum space and controlling the sample temperature for integrity are key challenges to the implementation of this technique. The break-thechain procedures for container configurations, which are being considered, were simulated by multi-physics finite element models. Different heating methods including induction and resistive/radiation were evaluated. The temperature profiles of Martian samples in a proposed container structure were predicted. The results show that the sealing and sterilizing process can be controlled such that the samples temperature is maintained below the level that may cause damage, and that the brazing technique is a feasible approach to breaking the contamination path.