WorldWideScience

Sample records for coolant loops

  1. Assessment of Candidate Molten Salt Coolants for the NGNP/NHI Heat-Transfer Loop

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D. F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2006-06-30

    This report provides an assessment of candidate salts proposed as the coolant for the loop that shuttles heat from the Next Generation Nuclear Plant (NGNP) to the Nuclear Hydrogen Initiative (NHI) hydrogen-production plant. The physical properties most relevant for coolant service were reviewed, and key chemical factors that influence material compatibility were also analyzed for the purpose of screening candidate salts. A preliminary assessment of the cost of the raw materials required to produce the coolant is also presented. Salts that are suitable for use as the primary coolant in a high-temperature nuclear reactor were previously analyzed. Some of the fluoride salts identified in the previous study are also appropriate for consideration as the secondary coolant in a heat-transfer loop; therefore, results from the previous report are used in this document. However, alternative coolant salts (i.e., chlorides and fluoroborates) that were not considered in the previous report should be considered for service in the heat-transfer loop. These alternative coolants are considered in this report.

  2. Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

  3. An Improved Design for Air Removal from Aerospace Fluid Loop Coolant Systems

    Science.gov (United States)

    Ritchie, Stephen M. C.; Holladay, Jon B.; Holt, J. Mike; Clark, Dallas W.

    2003-01-01

    Aerospace applications with requirements for large capacity heat removal (launch vehicles, platforms, payloads, etc.) typically utilize a liquid coolant fluid as a transport media to increase efficiency and flexibility in the vehicle design. An issue with these systems however, is susceptibility to the presence of noncondensable gas (NCG) or air. The presence of air in a coolant loop can have numerous negative consequences, including loss of centrifugal pump prime, interference with sensor readings, inhibition of heat transfer, and coolant blockage to remote systems. Hardware ground processing to remove this air is also cumbersome and time consuming which continuously drives recurring costs. Current systems for maintaining the system free of air are tailored and have demonstrated only moderate success. An obvious solution to these problems is the development and advancement of a passive gas removal device, or gas trap, that would be installed in the flight cooling system simplifying the initial coolant fill procedure and also maintaining the system during operations. The proposed device would utilize commercially available membranes thus increasing reliability and reducing cost while also addressing both current and anticipated applications. In addition, it maintains current pressure drop, water loss, and size restrictions while increasing tolerance for pressure increases due to gas build-up in the trap.

  4. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  5. Validation of computational fluid dynamics calculation using Rossendorf coolant mixing model flow measurements in primary loop of coolant in a pressurized water reactor model

    Energy Technology Data Exchange (ETDEWEB)

    Farkas, Istvan; Hutli, Ezddin; Faekas, Tatiana; Takacs, Antal; Guba, Attila; Toth, Ivan [Dept. of Thermohydraulics, Centre for Energy Research, Hungarian Academy of Sciences, Budapest (Hungary)

    2016-08-15

    The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

  6. Experimental study of electroinsulating coatings in gallium coolant related to the divertor cooling loop

    Science.gov (United States)

    Beznosov, A. V.; Sherbakov, R. V.; Karatushina, I. V.; Romanov, P. V.

    1996-10-01

    Experimental investigation of electroinsulating coatings stability on the samples made of stainless stell, vanadium alloy and beryllium has been conducted at 80-350°C. The impact of gas pressure upon the liquid gallium open surface was studied. The stability of electroinsulating film parameters on divertor structure materials was confirmed for the divertor with open liquid metal coolant surface in the vacuum chamber.

  7. Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

    Directory of Open Access Journals (Sweden)

    Avinash J. Gaikwad

    2008-01-01

    Full Text Available Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs, like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs is resorted to mitigate consequences of station blackout (SBO. In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT, SGs, and PDHRs under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections. On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

  8. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  9. Investigation of a hydrogen mitigation system during large break loss-of-coolant accident for a two-loop pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dehjourian, Mehdi; Rahgoshay, Mohmmad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

  10. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    ELC – Extended Life Coolant SCA – Supplemental Coolant Additive SOW – Scope of Work SwRI – Southwest Research Institute TARDEC – Tank Automotive...ethylene or propylene glycol and 35% extended life coolant #1 (ELC1) with a balance of water. At a higher ELC1 content of 45% or 50%, the mass loss...UNCLASSIFIED TABLE OF CONTENTS EXTENDED LIFE COOLANT TESTING INTERIM REPORT TFLRF No. 478 by Gregory A. T. Hansen Edwin A

  11. Machine coolant waste reduction by optimizing coolant life. Project summary

    Energy Technology Data Exchange (ETDEWEB)

    Pallansch, J.

    1995-08-01

    The project was designed to study the following: A specific water-soluble coolant (Blasocut 2000 Universal) in use with a variety of machines, tools, and materials; Coolant maintenance practices associated with three types of machines; Health effects of use and handling of recycled coolant; Handling practices for chips and waste coolant; Chip/coolant separation; and Oil/water separation.

  12. Technology of high-temperature organic coolant

    Energy Technology Data Exchange (ETDEWEB)

    Vorobei, M.P.; Makin, R.S.; Kuprienko, V.A. [and others

    1993-12-31

    A wide range of studies were carried out in RIAR on the problems connected with the use of high-temperature organic coolant at nuclear power plants. The work performed and successful experience gained in persistent operation of the ARBUS reactor confirmed the inherent safety characteristics, high operational reliability, as well as improved safety of stations with similar reactors. A large scope of studies were carried out at the ARBUS pilot reactor and loop with the organic coolant of the MIR reactor and a wide range of problems were solved. The studies are described.

  13. Environmentally Friendly Coolant System

    Energy Technology Data Exchange (ETDEWEB)

    David Jackson Principal Investigator

    2011-11-08

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  15. LMFBR with booster pump in pumping loop

    Science.gov (United States)

    Rubinstein, H.J.

    1975-10-14

    A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation.

  16. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-11-01

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

  18. Experience in operation of the experimental atomic power plant ''ARBUS'' with the high-boiling organic coolant-moderator ditolylmethane

    Energy Technology Data Exchange (ETDEWEB)

    Tzikanov, V.A.; Aleksenko, Yu.N.; Tetyukov, V.D.; Kuprienko, V.A.; Kobzar, I.G.; Khramchenkov, V.A.; Mexcheryakov, M.P.; Zinoviev, V.I.

    1978-04-01

    Radiolytic damage to the ditolylmethane organic coolant-moderator of the ARBUS reactor was removed by vacuum distillation. The majority of the degraded ditolylmethane formed gaseous and high-boiling materials, which were easily removed by the vacuum distillation. Unsaturated hydrocarbons and low-boiling residues were a minor contribution to the impurities produced by radiolysis in the primary coolant loop. Radioactivity in the primary coolant loop was found to be caused primarily from corrosion products of the system, /sup 16/N from dissolved oxygen, and impurities in the coolant-moderator. These also were significantly reduced in the vacuum distillation process.

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

    Science.gov (United States)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  1. Lithium as a blanket coolant

    Energy Technology Data Exchange (ETDEWEB)

    Wells, W.M.

    1977-01-01

    Recent re-assessment of tokamak reactors which move towards smaller size and lower required field strength (higher beta)/sup 2/ change the picture as regards the magnitude of MHD effects on flow resistance for lithium coolant. Perhaps the most important consequence of this as regards use of this coolant is that of clear acceptability of such effects when the flow is predominantly transverse to the magnetic field. This permits defining a blanket that consists entirely of round tubes containing the circulated lithium with voids between the tubes. Required thermal-hydraulic calculations are then on bases which are well established, especially in view of recent results dealing with perturbations of ducts and magnetic fields. Mitigation of MHD effects is feasible through tapering of tube wall thickness or use of insulated layers, but their use was not mandatory for the assumed conditions. Blanket configurations utilizing flowing lithium in round tubes immersed in static lithium may be suitable, but calculational methods do not now exist for this situation. Use of boiling potassium or cesium appears to be prohibitive in terms of vapor flow area when temperature levels are consistent with stainless steel. Liquid sodium, in addition to not being a breeding material, requires higher velocity than lithium for the same heat removal.

  2. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

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

  3. Technology of high temperature organic coolant

    Energy Technology Data Exchange (ETDEWEB)

    Makin, R.S.; Vorobei, M.P.; Kuprienko, V.A.; Starkov, V.A.; Tsykanov, V.A.; Checketkin, Y.V. [Research Institute of Atomic Reactors, Ulyanovsk (Russian Federation)

    1993-12-31

    Research has been performed on the problems related to the use of high temperature organic coolants in small and medium nuclear power plants. The work performed and also the experience of operating the ARBUS reactor confirmed the inherent safety features, reliability, and enhanced safety margins of the plants with this type of coolants. The advantages of this system and research highlights are presented.

  4. Flow boiling test of GDP replacement coolants

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.H. [comp.

    1995-08-01

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

  5. Computing Flows Of Coolants In Turbomachines

    Science.gov (United States)

    Meitner, P. L.

    1994-01-01

    Coolant Passage Flow (CPF) computer code developed to predict accurately coolant flow and heat transfer inside turbomachinery cooling passages (either radial or axial blading). Computes flow in one-inlet/one-outlet passage of any shape. Calculates rate of flow of coolant, temperature, pressure, velocity, and heat-transfer coefficients along passage. Integrates one-dimensional momentum and energy equations along defined flow path, taking into account change in area, addition or subtraction of mass, pumping, friction, and transfer of heat. Written in FORTRAN IV.

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

    NARCIS (Netherlands)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-30

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

  9. Mathematical Model-Based Temperature Preparation of Liquid-Propellant Components Cooled by Liquid Nitrogen in the Heat Exchanger with a Coolant

    Directory of Open Access Journals (Sweden)

    S. K. Pavlov

    2014-01-01

    Full Text Available Before fuelling the tanks of missiles, boosters, and spacecraft with liquid-propellant components (LPC their temperature preparation is needed. The missile-system ground equipment performs this operation during prelaunch processing of space-purpose missiles (SPM. Usually, the fuel cooling is necessary to increase its density and provide heat compensation during prelaunch operation of SPM. The fuel temperature control systems (FTCS using different principles of operation and types of coolants are applied for fuel cooling.To determine parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is cooled by liquid nitrogen upon contact heat exchange in the coolant reservoir, a mathematical model of this process and a design technique are necessary. Both allow us to determine design parameters of the cooling system and the required liquid nitrogen reserve to cool LPC to the appropriate temperature.The article presents an overview of foreign and domestic publications on cooling processes research and implementation using cryogenic products such as liquid nitrogen. The article draws a conclusion that it is necessary to determine the parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is liquid nitrogen-cooled upon contact heat exchange in the coolant reservoir allowing to define rational propellant cooling conditions to the specified temperature.The mathematical model describes the set task on the assumption that a heat exchange between the LPC and the coolant in the heat exchanger and with the environment through the walls of tanks and pipelines of circulation loops is quasi-stationary.The obtained curves allow us to calculate temperature changes of LPC and coolant, cooling time and liquid nitrogen consumption, depending on the process parameters such as a flow rate of liquid nitrogen, initial coolant temperature, pump characteristics, thermal

  10. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

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

  11. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Gene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-04-05

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  12. On-Line Coolant Chemistry Analysis

    Energy Technology Data Exchange (ETDEWEB)

    LM Bachman

    2006-07-19

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level.

  13. NGNP Reactor Coolant Chemistry Control Study

    Energy Technology Data Exchange (ETDEWEB)

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  14. Additional requirements for leak-before-break application to primary coolant piping in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G. [AIB Vincotte Nuclear, Brussels (Belgium)

    1997-04-01

    Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbers at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.

  15. Comparative Evaluation of Coolant Mixing Experiments at the ROCOM, Vattenfall, and Gidropress Test Facilities

    Directory of Open Access Journals (Sweden)

    S. Kliem

    2007-01-01

    Full Text Available Coolant mixing is an important mitigative mechanism against reactivity accidents caused by local boron dilution. Experiments on coolant mixing were carried out at three different test facilities representing three different reactor types. These are the ROCOM test facility modelling a German KONVOI-type reactor, the Vattenfall test facility being a model of a Westinghouse three-loop PWR, and the Gidropress test facility modelling a VVER-1000 PWR. The scenario of the start-up of the first main coolant pump was investigated in all three facilities. The experiments were accompanied by velocity measurements in the downcomer for the same scenario in the ROCOM and the Vattenfall test facilities. A similar flow structure was found in these measurements in both cases. A maximum of the velocity is measured at the opposite side in regard to the position of the loop with the starting-up pump whilst a recirculation area was found just below this inlet nozzle in both facilities. The analysis of the slug mixing experiments showed also comparable flow behaviour. In accordance with the velocity measurements, the first part of the deboration is also found on the opposite side. In this region, the maximum deboration is measured in all three cases. These maximum values are in the same order of magnitude for nearly identical initial slug volumes.

  16. Recovery studies for plutonium machining oil coolant

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-04-27

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

  17. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

  18. Corrosion of magnesium alloys in commercial engine coolants

    Energy Technology Data Exchange (ETDEWEB)

    Song, G.; StJohn, D.H. [CRC for Cast Metals Manufacturing (CAST), Division of Materials, School of Engineering, The University of Queensland, Brisbane, QLD 4072 (Australia)

    2005-01-01

    A number of magnesium alloys show promise as engine block materials. However, a critical issue for the automotive industry is corrosion of the engine block by the coolant and this could limit the use of magnesium engine blocks. This work assesses the corrosion performance of conventional magnesium alloy AZ91D and a recently developed engine block magnesium alloy AM-SC1 in several commercial coolants. Immersion testing, hydrogen evolution measurement, galvanic current monitoring and the standard ASTM D1384 test were employed to reveal the corrosion performance of the magnesium alloys subjected to the coolants. The results show that the tested commercial coolants are corrosive to the magnesium alloys in terms of general and galvanic corrosion. The two magnesium alloys exhibited slightly different corrosion resistance to the coolants with AZ91D being more corrosion resistant than AM-SC1. The corrosivity varied from coolant to coolant. Generally speaking, an organic-acid based long life coolant was less corrosive to the magnesium alloys than a traditional coolant. Among the studied commercial coolants, Toyota long life coolant appeared to be the most promising one. In addition, it was found that potassium fluoride effectively inhibited corrosion of the magnesium alloys in the studied commercial coolants. Both general and galvanic corrosion rates were significantly decreased by addition of KF, and there were no evident side effects on the other engine block materials, such as copper, solder, brass, steel and aluminium alloys, in terms of their corrosion performance. The ASTM D 1384 test further confirmed these results and suggested that Toyota long life coolant with 1%wt KF addition is a promising coolant for magnesium engine blocks. (Abstract Copyright [2005], Wiley Periodicals, Inc.)

  19. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A., E-mail: anlafuente@etsii.upm.e [ETSII-UPM, c/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Piera, M. [ETSII:UNED, c/Juan del Rosal, 12, 28040 Madrid (Spain)

    2010-06-15

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F{sub 2}Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F{sub 2}Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void

  20. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A. [ETSI Industriales-Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal, 2. 28006 Madrid (Spain)

    2010-07-01

    Breeder reactors are considered the unique tool for fully exploiting the natural nuclear resources. In current LWR, only a 0.5% of the primary energy contained in the nuclei removed from the mine is converted into useful heat, with the rest remaining in the depleted uranium or in the spent fuel. The objective of resource-efficiency stimulated the interest in Fast- Reactor-based fuel cycles which can exploit a much higher fraction of the energy content of the mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles would also offers several potential advantages over a uranium fuel cycle. The coolant initially chosen for most of the FBR programs launched in the 60's was sodium, which still is considered the best candidate for these reactors. However, Na-cooled FBR have a positive void reactivity coefficient, which has been among others, a fundamental drawback that has cancelled the deployment of these reactors. Therefore, it seems reasonable to explore totally new options on coolants for breeders. In this paper, a proposal is presented on a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would need an extensive R and D programme, this paper presents the very appealing properties of this salt, in the case of using a specific type of fuel, similar to that of pebble bed reactors. The concept will be studied over a typical MOX composition and extended to a Thorium-based cycle. The general analysis takes into account requirements for criticality (opening the option of hybrid subcritical systems); requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window is found in the definition of a F{sub 2}Be cooled reactor where the safety requirement is met, unlike for molten metal cooled reactors which always have positive void

  1. Assessment and Accommodation of Thermal Expansion of the Internal Active Thermal Control System Coolant During Launch to On-Orbit Activation of International Space Station Elements

    Science.gov (United States)

    Edwards, Darryl; Ungar, Eugene K.; Holt, James M.

    2002-01-01

    The International Space Station (ISS) employs an Internal Active Thermal Control System (IATCS) comprised of several single-phase water coolant loops. These coolant loops are distributed throughout the ISS pressurized elements. The primary element coolant loops (i.e. U.S. Laboratory module) contain a fluid accumulator to accomodate thermal expansion of the system. Other element coolant loops are parasitic (i.e. Airlock), have no accumulator, and require an alternative approach to insure that the system maximum design pressure (MDP) is not exceeded during the Launch to Activation (LTA) phase. During this time the element loops is a stand alone closed system. The solution approach for accomodating thermal expansion was affected by interactions of system components and their particular limitations. The mathematical solution approach was challenged by the presence of certain unknown or not readily obtainable physical and thermodynamic characteristics of some system components and processes. The purpose of this paper is to provide a brief description of a few of the solutions that evolved over time, a novel mathematical solution to eliminate some of the unknowns or derive the unknowns experimentally, and the testing and methods undertaken.

  2. Rotor dynamic analysis of main coolant pump

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chong Won; Seo, Jeong Hwan; Kim, Choong Hwan; Shin, Jae Chul; Wang, Lei Tian [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    A rotor dynamic analysis program DARBS/MCP, for the main coolant pump of the integral reactor, has been developed. The dynamic analysis model of the main coolant pump includes a vertical shaft, three grooved radial journal bearings and gaps that represent the structure-fluid interaction effects between the rotor and the lubricant fluid. The electromagnetic force from the motor and the hydro-dynamic force induced by impeller are the major sources of vibration that may affect the rotor system stability. DARBS/MCP is a software that is developed to effectively analyze the dynamics of MCP rotor systems effectively by applying powerful numerical algorithms such as FEM with modal truncation and {lambda}-matrix method for harmonic analysis. Main design control parameters, that have much influence to the dynamic stability, have been found by Taguchi's sensitivity analysis method. Design suggestions to improve the stability of MCP rotor system have been documented. The dynamic bearing parameters of the journal bearings used for main coolant pump have been determined by directly solving the Reynolds equation using FDM method. Fluid-structure interaction effect that occurs at the small gaps between the rotor and the stator were modeled as equivalent seals, the electromagnetic force effect was regarded as a linear negative radial spring and the impeller was modeled as a rigid disk with hydrodynamic and static radial force. Although there exist critical speeds in the range of operational speeds for type I and II rotor systems, the amplitude of vibration appears to be less than the vibration limit set by the API standards. Further more, it has been verified that the main design parameters such as the clearance and length of journal bearings, and the static radial force of impeller should be properly adjusted, in order to the improve dynamic stability of the rotor system. (author). 39 refs., 81 figs., 17 tabs.

  3. Cryogenic-coolant He-4-superconductor interaction

    Science.gov (United States)

    Caspi, S.; Lee, J. Y.; Kim, Y. I.; Allen, R. J.; Frederking, T. H. K.

    1978-01-01

    The thermodynamic and thermal interaction between a type 2 composite alloy and cryo-coolant He4 was studied with emphasis on post quench phenomena of formvar coated conductors. The latter were investigated using a heater simulation technique. Overall heat transfer coefficients were evaluated for the quench onset point. Heat flux densities were determined for phenomena of thermal switching between a peak and a recovery value. The study covered near saturated liquid, pressurized He4, both above and below the lambda transition, and above and below the thermodynamic critical pressure. In addition, friction coefficients for relative motion between formvar insulated conductors were determined.

  4. Liquid metal reactor development -Studies on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

  5. A study on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  6. Modeling of melt-coolant mixing by bottom injection

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkov, I.V.; Paladino, D.; Sehgal, B.R. [Royal Inst. of Tech., Div. of Nuclear Power Safety, Stockholm (Sweden)

    2001-07-01

    In this paper, the flow characteristics during the coolant injection, with submerged nozzles, at the bottom of a molten pool are studied. The flow pattern developed by the rising coolant is considered for the case of complete coolant vaporization, and the pool-coolant phase distributions are assessed by a modeling approach delivered from literature for a heterogeneous turbulent jet. To calculate the basic characteristics of such flow, integral relationships are proposed for the two-phase boundary layer. The results of numerical computations and approximate solution are compared with the experimental data obtained in the low temperature experiments, conducted in the DECOBI (debris coolability by bottom injection) facility. (authors)

  7. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  8. Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

    2009-06-15

    would be important for NPP life time management purposes. In a similar way it is possible to lead estimation of EFCPO, Q - factors of coolant acoustic oscillatory circuit and PBF for any of updating NPP with PWR including NPP with supercritical parameters. Certainly, the quantitative characteristics of EFCPO, Q - factors and PBF will be various for each class of the nuclear reactor. Paper shows what operating control influences are necessary to remove EFCPO from area of resonant interaction with vibrations FR, FA etc. It is offered to use instrumentation and control systems to prevent operating of NPP at capacity level which provides increasing in amplitudes of pulsations of pressure. The increase in demand of the safety of NPP requires further increase of adequacy between a model and an object. The integrated PSB-VVER test facility is the 1:300 replica of the prototype reactor VVER with respect to power capacity and volume. The height evaluations of the test facility are the same as those of the original. The maximum power of heat released by an assembly of fuel rod simulators is 10 MW. PSB-VVER consists of four loops closed to the reactor model; the latter consists of a down comer section with the lower mixing chamber, a model of the reactor core (a channel with fuel rod simulators), a bypass of the reactor core model, and the upper mixing chamber. Each loop contains a reactor coolant pump, a steam generator, and a cold and hot pipeline. The test facility also includes a pressurizer and an ECCS consisting of three subsystems: a passive one, which incorporates four hydro accumulators and two active ones (a high-pressure ECCS and a low pressure ECCS). Test facility description, scheme and the measuring system are presented. Using such systems the transient processes have been investigated in accident with loss of coolant from the primary cooling system. The basic mathematical models for calculation of EFCPO are achieved. These models are intended for both one-phase and

  9. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Eugene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-03-22

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  10. Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

    Energy Technology Data Exchange (ETDEWEB)

    Soli T. Khericha

    2006-09-01

    This report presents preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T&FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420oC. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  12. Power module assemblies with staggered coolant channels

    Science.gov (United States)

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

    2013-07-16

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Yamano, Norihiko; Maruyama, Yu; Moriyama, Kiyofumi; Yang, Y.; Sugimoto, Jun [Severe Accident Research Laboratory, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    1999-07-01

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

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

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

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

  15. SUBSTATIONS OF DISTRICT HEATING SYSTEMS WITH PULSE COOLANT CIRCULATION

    Directory of Open Access Journals (Sweden)

    Andrey N. Makeev

    2017-01-01

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

  16. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  17. Analysis of a hypothetical loss of coolant accident in a Konvoi type NPP by GASFLOW and COCOSYS

    Energy Technology Data Exchange (ETDEWEB)

    Benz, Stefan; Royl, Peter [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Band, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    2013-07-01

    The 3D computational fluid dynamics code GASFLOW and the German containment code system COCOSYS, which is based on a lumped-parameter approach, are used to simulate the hydrogen-air-steam distribution and hydrogen mitigation in a Konvoi type nuclear power plant in a postulated hypothetical core melt accident. A break in a coolant loop and the subsequent loss of the coolant causes a strong heat-up of the core. As a consequence hydrogen is produced by oxidation of cladding tubes. The residual steam and the produced hydrogen are released into the containment through the break in the coolant loop. Without suitable counter measures, sensitive mixtures can build up with a combustion potential which could threaten the integrity of the containment. A model of a Konvoi type nuclear power plant which is equipped with passive autocatalytic recombiners is used to simulate such accident scenario. COCOSYS allows comprehensive simulation of all relevant processes of severe accidents, whereas GASFLOW is primarily designed to simulate the distribution of steam and hydrogen within the containment. This paper presents the comparison of GASFLOW and COCOSYS simulation results for the in-vessel phase of the selected accident. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  19. Corrosion problems with aqueous coolants, final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-04-11

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

  20. Lubricant-coolant fluid for machining metals

    Energy Technology Data Exchange (ETDEWEB)

    Berlin, A.A.; Epshtein, V.R.; Pastunov, V.A.; Sherle, A.I.; Shpin' kov, V.A.; Sladkova, T.A.

    1981-03-10

    For improving the antiwear and anticorrosion properties, the lubricant-coolant fluid (LCF) based on water, triethanolamine, and NaNO/sub 2/ contains additionally the sodium salt of an acid ester of maleic acid and substituted oligooxyethylenes (NMO) with the following proportions of the components: triethanolamine 0.3-0.5%, NaNO/sub 2/ 0.3-0.5%, NMO 0.5-2.0%, and water the remainder. In the case of using the proposed LCF on high-speed machine tools, it can contain additionally a foam suppressor in an amount of 0.005-0.1%. For preventing microbiological contamination of the LCF, bactericides of the type furacillin, formalin, vazin (transliteration), and others in an amount of 0.005-0.1% can be added to its composition. Introduction of the NMO additive ensures high wetting and lubricating characteristics in the LCF, which is characterized by stability during storage and service and good anticorrosion properties. Use of the proposed LCF makes it possible to increase the life of the cutting tool by a factor of 2.2 in machining Steel 40Kh and by a factor of 1.3 in machining corroding steel by comparison with the prototype; at the same time the service life of the LCF is increased twofold. The LCF can be used in machining parts of alloyed construction and corrosionresistant steels with cutting-edge and abrasive tools.

  1. Transient two-phase performance of LOFT reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  3. Optimized Coolant-Flow Diverter For Increased Bearing Life

    Science.gov (United States)

    Subbaraman, Maria R.; Butner, Myles F.

    1995-01-01

    Coolant-flow diverter for rolling-element bearings in cryogenic turbopump designed to enhance cooling power of flow in contact with bearings and thereby reduce bearing wear. Delivers jets of coolant as close as possible to hot spots at points of contact between balls and race. Also imparts swirl that enhances beneficial pumping effect. Used with success in end ball bearing of high-pressure-oxidizer turbopump.

  4. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States). Dept. of Mechanical Engineering

    1995-12-31

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines; however there is practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

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

    Science.gov (United States)

    2006-12-01

    1992) PFR UK 250 MWe - 14 Shut Down (1994) Rapsodie France 40 MWe - 40 Shut Down (1983) Phenix France 233 MWe - 22 In Operation BOR-60 Russia...107years.98 • Problems with radioactive waste management and coolant disposal during decommissioning .99 O th er • Lead is abundantly available in...is high due to Bi-210, half-life 3.6 106years.102 • Problems with radioactive waste management and coolant disposal during decommissioning . 103 O

  6. Analysis of boron dilution in a four-loop PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sun, J.G.; Sha, W.T. [Argonne National Lab., IL (United States)

    1995-03-01

    Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four-loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant is at hot standby and the reactor coolant system has been heated to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated and a sensitivity study is performed to assess the accuracy of the numerical modeling of the geometry of the reactor coolant system.

  7. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  8. Integrated Vehicle Thermal Management - Combining Fluid Loops in Electric Drive Vehicles (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Rugh, J. P.

    2013-07-01

    Plug-in hybrid electric vehicles and electric vehicles have increased vehicle thermal management complexity, using separate coolant loop for advanced power electronics and electric motors. Additional thermal components result in higher costs. Multiple cooling loops lead to reduced range due to increased weight. Energy is required to meet thermal requirements. This presentation for the 2013 Annual Merit Review discusses integrated vehicle thermal management by combining fluid loops in electric drive vehicles.

  9. Use of microPCM fluids as enhanced liquid coolants in automotive EV and HEV vehicles. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mulligan, James C.; Gould, Richard D.

    2001-10-31

    Proof-of-concept experiments using a specific microPCM fluid that potentially can have an impact on the thermal management of automotive EV and HEV systems have been conducted. Samples of nominally 20-micron diameter microencapsulated octacosane and glycol/water coolant were prepared for testing. The melting/freezing characteristics of the fluid, as well as the viscosity, were determined. A bench scale pumped-loop thermal system was used to determine heat transfer coefficients and wall temperatures in the source heat exchanged. Comparisons were made which illustrate the enhancements of thermal performance, reductions of pumping power, and increases of heat transfer which occur with the microPCM fluid.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

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

  12. Actively controlling coolant-cooled cold plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  13. Experimental study of high temperature particle dropping in coolant liquid

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; LI Xiaoyan; HU Zhihua

    2007-01-01

    A series of experiments of the premixing stage of fuel-coolant interactions (FCI), namely the particles falling into water, were carried out. The force on the particles during the course of falling has been studied. The dropping character of hot particle was influenced by three main parameters, i.e., particle temperature, particle diameter and coolant subcooling that varied over a wide range. A high-speed camera recorded the falling speed of the particle and the moving curves were obtained. The experimental results showed that for the film boiling on the surface of particle and water, the temperature increase of either particle or coolant would slow down the particle falling velocity. The falling velocity of particle in small diameter is lower than that of the bigger particle. The present work can provide an experimental foundation for further investigation of high-speed transient evaporation heat transfer.

  14. Fuel cell cooling system using a non-dielectric coolant

    Energy Technology Data Exchange (ETDEWEB)

    Grevstad, P.E.; Gelting, R.L.

    1976-07-13

    A cooler for removing waste heat from a stack of fuel cells uses a non-dielectric coolant which is carried in a plurality of tubes passing through one or more separator plates in the stack. Preferably the coolant is water so that heat removal is by evaporation of the water within the tubes by boiling. The tubes are electrically insulated from the cells by a coating of dielectric material such as polytetrafluoroethylene. In one embodiment of the invention the cooler tubes are connected to the stack coolant supply conduits by dielectric hoses having a high length to diameter ratio to provide a several hundred thousand ohm impedance path in case of a flaw in the protective dielectric coating, in order that a short circuit of the stack does not occur.

  15. Investigation on two-phase critical flow for loss-of-coolant accident of pressurized water reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    The previous investigations were mainly conducted under the condition of low pressure,however,the steam-water specific volume and the interphase evaporation rate in high pressure are much different from those in low pressure,Therefore,the new experimental and theoretical investigation are performed in Xi'an Jiaotong University.The investigation results could be directly applied to the analysis of loss-of -coolant accident for pressurized water reacor.The system transition characteristics of cold leg and hot leg break loss-of -coolant tests are described for convective circulation test loop.Two types of loss-of-coolant accident are identified for :hot leg” break,while three types for “cold leg”break and the effect parameters on the break geometries.Tests indicate that the mass flow rate with convergent-divergent nozzle reaches the maximum value among the different break sections at the same inlet fluid condition because the fluid separation does not occur.A wall surface cavity nucleation model is developed for prediction of the critical mass flow rate with water flowing in convergentdivergent nozzles.

  16. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  17. Reactor coolant pump shaft seal behavior during blackout conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue.

  18. Simulation of a transient with loss of primary coolant due to a small rupture in Angra 2 nuclear power plant with RELAP5/MOD3.2.2G code; Simulacao de um acidente postulado de perda de refrigerante primario por pequena ruptura na usina de Angra 2 com o codigo RELAP5/MOD3.2.2G

    Energy Technology Data Exchange (ETDEWEB)

    Sabundjian, Gaiane; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    This paper presents a nodalization for Angra 2 Nuclear Power Plant, as well as the results obtained for a Small Break Loss of Coolant Accident (SBLOCA), simulated with RELAP5/MOD3.2G code. This accident consists in a small break (380 m{sup 2}) in the line of the Emergency Core Coolant System (ECCS) in loop 20 of Angra 2. Results are not as expected, however they are satisfactory regarding the nodalization used. (author)

  19. Analysis of boron dilution in a four-loop PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sun, J.G.; Sha, W.T.

    1995-12-31

    Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant is at hot standby and the reactor coolant system has been heated up to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated.

  20. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY DISTILLATION

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants for a facility such as the New Jersey Department of Transportation garage in Ewing, New Jersey. he specific recycling evaluated is b...

  1. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT.

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants at a New Jersey Department of Transportation garage. The specific recycling units evaluated are based on the technologies of filtrat...

  2. Fuels, Lubricants, and Coolants. FOS: Fundamentals of Service.

    Science.gov (United States)

    John Deere Co., Moline, IL.

    This manual on fuels, lubricants, and coolants is one of a series of power mechanics tests and visual aids on automotive and off-the-road agricultural and construction equipment. Materials present basic information with illustrations for use by vocational students and teachers as well as shop servicemen and laymen. Focusing on fuels, the first of…

  3. Integral coolant channels supply made by melt-out method

    Science.gov (United States)

    Escher, W. J. D.

    1964-01-01

    Melt-out method of constructing strong, pressure-tight fluid coolant channels for chambers is accomplished by cementing pins to the surface and by depositing a melt-out material on the surface followed by two layers of epoxy-resin impregnated glass fibers. The structure is heated to melt out the low-melting alloy.

  4. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT.

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants at a New Jersey Department of Transportation garage. The specific recycling units evaluated are based on the technologies of filtrat...

  5. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY DISTILLATION

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants for a facility such as the New Jersey Department of Transportation garage in Ewing, New Jersey. he specific recycling evaluated is b...

  6. Pump, and earth-testable spacecraft capillary heat transport loop using augmentation pump and check valves

    Science.gov (United States)

    Baker, David (Inventor)

    1998-01-01

    A spacecraft includes heat-generating payload equipment, and a heat transport system with a cold plate thermally coupled to the equipment and a capillary-wick evaporator, for evaporating coolant liquid to cool the equipment. The coolant vapor is coupled to a condenser and in a loop back to the evaporator. A heated coolant reservoir is coupled to the loop for pressure control. If the wick is not wetted, heat transfer will not begin or continue. A pair of check valves are coupled in the loop, and the heater is cycled for augmentation pumping of coolant to and from the reservoir. This augmentation pumping, in conjunction with the check valves, wets the wick. The wick liquid storage capacity allows the augmentation pump to provide continuous pulsed liquid flow to assure continuous vapor transport and a continuously operating heat transport system. The check valves are of the ball type to assure maximum reliability. However, any type of check valve can be used, including designs which are preloaded in the closed position. The check valve may use any ball or poppet material which resists corrosion. For optimum performance during testing on Earth, the ball or poppet would have neutral buoyancy or be configured in a closed position when the heat transport system is not operating. The ball may be porous to allow passage of coolant vapor.

  7. Loop-to-loop coupling.

    Energy Technology Data Exchange (ETDEWEB)

    Warne, Larry Kevin; Lucero, Larry Martin; Langston, William L.; Salazar, Robert Austin; Coleman, Phillip Dale; Basilio, Lorena I.; Bacon, Larry Donald

    2012-05-01

    This report estimates inductively-coupled energy to a low-impedance load in a loop-to-loop arrangement. Both analytical models and full-wave numerical simulations are used and the resulting fields, coupled powers and energies are compared. The energies are simply estimated from the coupled powers through approximations to the energy theorem. The transmitter loop is taken to be either a circular geometry or a rectangular-loop (stripline-type) geometry that was used in an experimental setup. Simple magnetic field models are constructed and used to estimate the mutual inductance to the receiving loop, which is taken to be circular with one or several turns. Circuit elements are estimated and used to determine the coupled current and power (an equivalent antenna picture is also given). These results are compared to an electromagnetic simulation of the transmitter geometry. Simple approximate relations are also given to estimate coupled energy from the power. The effect of additional loads in the form of attached leads, forming transmission lines, are considered. The results are summarized in a set of susceptibility-type curves. Finally, we also consider drives to the cables themselves and the resulting common-to-differential mode currents in the load.

  8. Improvements of primary coolant shutdown chemistry and reactor coolant system cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Gaudard, G.; Gilles, B.; Mesnage, F. [EDF/GDL (France); Cattant, F. [EDF R and D (France)

    2002-07-01

    In the framework of a radiation exposure management program entitled <>, EDF aims at decreasing the mass dosimetry of nuclear power plants workers. So, the annual dose per unit, which has improved from 2.44 m.Sv in 1991 to 1.08 in 2000, should target 0.8 mSv in the year 2005 term in order to meet the results of the best nuclear operators. One of the guidelines for irradiation source term reduction is the optimization of operation parameters, including reactor coolant system (RCS) chemistry in operation, RCS shutdown chemistry and RCS cleanup improvement. This paper presents the EDF strategy for the shutdown and start up RCS chemistry optimization. All the shutdown modes have been reviewed and for each of them, the chemical specifications will be fine tuned. A survey of some US PWRs shutdown practices has been conducted for an acid and reducing shutdown chemistry implementation test at one EDF unit. This survey shows that deviating from the EPRI recommended practice for acid and reducing shutdown chemistry is possible and that critical path impact can be minimized. The paper also presents some investigations about soluble and insoluble species behavior and characterization; the study focuses here on {sup 110m}Ag, {sup 122}Sb, {sup 124}Sb and iodine contamination. Concerning RCS cleanup improvement, the paper presents two studies. The first one highlights some limited design modifications that are either underway or planned, for an increased flow rate during the most critical periods of the shutdown. The second one focuses on the strategy EDF envisions for filters and resins selection criteria. Matching the study on contaminants behavior with the study of filters and resins selection criteria should allow improving the cleanup efficiency. (authors)

  9. Surface Treatment to Improve Corrosion Resistance in Lead-Alloy Coolants

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen; Kumar Sridharan; McLean T. Machut; Lizhen Tan

    2007-08-29

    One of the six proposed advanced reactor designs of the Generation IV Initiative, the Leadcooled Fast Reactor (LFR) possesses many characteristics that make it a desirable candidate for future nuclear energy production and responsible actinide management. These characteristics include favorable heat transfer, fluid dynamics, and neutronic performance compared to other candidate coolants. However, the use of a heavy liquid metal coolant presents a challenge for reactor designers in regards to reliable structural and fuel cladding materials in both a highly corrosive high temperature liquid metal and an intense radiation fieldi. Flow corrosion studies at the University of Wisconsin have examined the corrosion performance of candidate materials for application in the LFR concept as well as the viability of various surface treatments to improve the materials’ compatibility. To date this research has included several focus areas, which include the formulation of an understanding of corrosion mechanisms and the examination of the effects of chemical and mechanical surface modifications on the materials’ performance in liquid lead-bismuth by experimental testing in Los Alamos National Laboratory’s DELTA Loop, as well as comparison of experimental findings to numerical and physical models for long term corrosion prediction. This report will first review the literature and introduce the experiments and data that will be used to benchmark theoretical calculations. The experimental results will be followed by a brief review of the underlying theory and methodology for the physical and theoretical models. Finally, the results of theoretical calculations as well as experimentally obtained benchmarks and comparisons to the literature are presented.

  10. Hybrid method for numerical modelling of LWR coolant chemistry

    Science.gov (United States)

    Swiatla-Wojcik, Dorota

    2016-10-01

    A comprehensive approach is proposed to model radiation chemistry of the cooling water under exposure to neutron and gamma radiation at 300 °C. It covers diffusion-kinetic processes in radiation tracks and secondary reactions in the bulk coolant. Steady-state concentrations of the radiolytic products have been assessed based on the simulated time dependent concentration profiles. The principal reactions contributing to the formation of H2, O2 and H2O2 were indicated. Simulation was carried out depending on the amount of extra hydrogen dissolved in the coolant to reduce concentration of corrosive agents. High sensitivity to the rate of reaction H+H2O=OH+H2 is shown and discussed.

  11. Expert system for online surveillance of nuclear reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-12-31

    This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  12. Effect of coolant inhibitors on AZ91D

    Institute of Scientific and Technical Information of China (English)

    I.M. Baghni; WU Yinshun; ZHANG Wei; LI Jiuqing

    2004-01-01

    The inhibition effects of sodium vanadate along with inorganic coolant inhibitors were examined on corrosion of AZ91D in ASTM D1384-80 corrosive water by polarization measurements. The galvanic corrosion of AZ91D coupled to 3003, 6063, and 356 Al alloys were also tested. An effective combination of inhibitors containing (but not limited to) sodium vanadate, silicate, and nitrate was proposed for inhibition of AZ91D and prevention of galvanic corrosion.

  13. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  14. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  15. Post-irradiation data on fuel elements from KER Loop 4

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, E.C.

    1963-01-10

    Fourteen NAE1 fuel elements were discharged from KER Loop-4, after irradiation to an average exposure of 1250 MWD, at prototype N-Reactor coolant temperature and pressure. The elements were disassembled and measured in the KE fuel examination facility. This report includes all measurements, except the profilometer data.

  16. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  17. Effect of Coolant Temperature and Mass Flow on Film Cooling of Turbine Blades

    Science.gov (United States)

    Garg, Vijay K.; Gaugler, Raymond E.

    1997-01-01

    A three-dimensional Navier Stokes code has been used to study the effect of coolant temperature, and coolant to mainstream mass flow ratio on the adiabatic effectiveness of a film-cooled turbine blade. The blade chosen is the VKI rotor with six rows of cooling holes including three rows on the shower head. The mainstream is akin to that under real engine conditions with stagnation temperature = 1900 K and stagnation pressure = 3 MPa. Generally, the adiabatic effectiveness is lower for a higher coolant temperature due to nonlinear effects via the compressibility of air. However, over the suction side of shower-head holes, the effectiveness is higher for a higher coolant temperature than that for a lower coolant temperature when the coolant to mainstream mass flow ratio is 5% or more. For a fixed coolant temperature, the effectiveness passes through a minima on the suction side of shower-head holes as the coolant to mainstream mass flow, ratio increases, while on the pressure side of shower-head holes, the effectiveness decreases with increase in coolant mass flow due to coolant jet lift-off. In all cases, the adiabatic effectiveness is highly three-dimensional.

  18. System Study: High-Pressure Coolant Injection 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. 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 or decreasing trends were identified in the HPCI results.

  19. Reactor coolant pump shaft seal stability during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, D B; Hill, R C; Wensel, R G

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries.

  20. System Study: High-Pressure Coolant Injection 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 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 or decreasing trends were identified in the HPCI results.

  1. Simulation of 3D Flow in Turbine Blade Rows including the Effects of Coolant Ejection

    Institute of Scientific and Technical Information of China (English)

    Jian-Jun LIU; Bai-Tao AN; Yun-Tao ZENG

    2008-01-01

    This paper describes the numerical simulation of three-dimensional viscous flows in air-cooled turbine blade rows with the effects of coolant ejection. A TVD Navier-Stokes flow solver incorporated with Baldwin-Lomax turbulence model and multi-grid convergence acceleration algorithm are used for the simulation. The influences of coolant ejection on the main flow are accounted by volumetric coolant source terms. Numerical results for a four-stage turbine are presented and discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  3. Uniform corrosion of FeCrAl alloys in LWR coolant environments

    Science.gov (United States)

    Terrani, K. A.; Pint, B. A.; Kim, Y.-J.; Unocic, K. A.; Yang, Y.; Silva, C. M.; Meyer, H. M.; Rebak, R. B.

    2016-10-01

    The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300-500 μm thick cladding.

  4. Simulating experimental investigation on the safety of nuclear heating reactor in loss-of-coolant accidents

    Science.gov (United States)

    Xu, Zhanjie

    1996-12-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. Up to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as follos: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  5. Station blackout with reactor coolant pump seal leakage

    Energy Technology Data Exchange (ETDEWEB)

    Evinay, A. (Southern California Edison, Irvine, CA (United States))

    1993-01-01

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

  6. Aqueous Nanofluid as a Two-Phase Coolant for PWR

    Directory of Open Access Journals (Sweden)

    Pavel N. Alekseev

    2012-01-01

    Full Text Available Density fluctuations in liquid water consist of two topological kinds of instant molecular clusters. The dense ones have helical hydrogen bonds and the nondense ones are tetrahedral clusters with ice-like hydrogen bonds of water molecules. Helical ordering of protons in the dense water clusters can participate in coherent vibrations. The ramified interface of such incompatible structural elements induces clustering impurities in any aqueous solution. These additives can enhance a heat transfer of water as a two-phase coolant for PWR due to natural forming of nanoparticles with a thermal conductivity higher than water. The aqueous nanofluid as a new condensed matter has a great potential for cooling applications. It is a mixture of liquid water and dispersed phase of extremely fine quasi-solid particles usually less than 50 nm in size with the high thermal conductivity. An alternative approach is the formation of gaseous (oxygen or hydrogen nanoparticles in density fluctuations of water. It is possible to obtain stable nanobubbles that can considerably exceed the molecular solubility of oxygen (hydrogen in water. Such a nanofluid can convert the liquid water in the nonstoichiometric state and change its reduction-oxidation (RedOx potential similarly to adding oxidants (or antioxidants for applying 2D water chemistry to aqueous coolant.

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

    Science.gov (United States)

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

    2008-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, R.K.; Sinha, S.K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Madhusoodanan, K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: kmadhu@barc.gov.in

    2008-12-15

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

  9. SIMMER-III Analyses of Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool: Effect of Coolant Quantity

    Directory of Open Access Journals (Sweden)

    Songbai Cheng

    2015-01-01

    Full Text Available Studies on local fuel-coolant interactions (FCI in a molten pool are important for the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs. To clarify the mechanisms underlying this interaction, in recent years, several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.

  10. Alternative loop rings

    CERN Document Server

    Goodaire, EG; Polcino Milies, C

    1996-01-01

    For the past ten years, alternative loop rings have intrigued mathematicians from a wide cross-section of modern algebra. As a consequence, the theory of alternative loop rings has grown tremendously. One of the main developments is the complete characterization of loops which have an alternative but not associative, loop ring. Furthermore, there is a very close relationship between the algebraic structures of loop rings and of group rings over 2-groups. Another major topic of research is the study of the unit loop of the integral loop ring. Here the interaction between loop rings and group ri

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

    Directory of Open Access Journals (Sweden)

    N. Dharmaraju

    2008-01-01

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

  12. Nonflammable coolants for space vehicle environmental control systems Compatibility of component materials with selected dielectric fluids.

    Science.gov (United States)

    Howard, R. T.; Korpolinski, T. S.; Mace, E. W.

    1971-01-01

    This paper summarizes a 4-year effort to evaluate and implement a nonflammable substitute coolant for application in the Saturn instrument unit (IU) environmental control system (ECS). Discussed are candidate material evaluations, detailed investigations of the properties of the coolant selected, and a summary of the implementation into a flight vehicle.

  13. Reactor coolant pump shaft seal behavior during station blackout

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-04-01

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

  14. Actively controlling coolant-cooled cold plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Parida, Pritish R.

    2016-04-26

    Cooling apparatuses are provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The cooling apparatus includes the cold plate and a controller. The cold plate couples to one or more electronic components to be cooled, and includes an adjustable physical configuration. The controller dynamically varies the adjustable physical configuration of the cold plate based on a monitored variable associated with the cold plate or the electronic component(s) being cooled by the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the electronic component(s), and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the cold plate, the positioning of which may be adjusted based on the monitored variable.

  15. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs.

  16. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-02

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  19. Hydrodynamics of heavy liquid metal coolant processes and filtering apparatus

    Energy Technology Data Exchange (ETDEWEB)

    Albert K Papovyants; Yuri I Orlov; Pyotr N Martynov; Yuri D Boltoev [Institute for Physics and Power Engineering named after A.I. Leypunsky Bondarenko sq. 1, 249033, Obninsk, Kaluga region (Russian Federation)

    2005-07-01

    Full text of publication follows: To optimize the design of filters for cleaning heavy liquid metal coolant (HLMC) from suspended impurities and choose appropriate filter material, the contribution is considered of different mechanisms of delivery and retention of these impurities from the coolant flow, which is governed by its specificity as a thermodynamically instable disperse system to a large extent. It is shown that the buildup of deposits in the filter is favored by the hydrodynamic regime with minimum filtration rates being due to the predominance in the suspension of the fine-dispersed solid phase (oxides Fe{sub 3}O{sub 4}, Cr{sub 2}O{sub 3} and so on). With concentrating the last mentioned phase in filter material pores or stagnant zones, coagulation structuration is possible, which is accompanied by sharp local increase in the viscosity and strength of the solid phase medium being built from liquid metal, i.e. slag sedimentary deposits. In rather extended pores, disintegration of such structures is possible, which is accompanied by sedimentation of large particles produced due to sticking together at coagulation. The analytical solution of the problem of particle sedimentation due to diffusion indicated that in the case under consideration, this mechanism takes place for particles less than {approx} 0,05 {mu}m in size, which is specified by the fact that the time of their delivery to the filter material surface is longer than that of the coolant being in the filter. The London-Van-der-Waals molecular forces play a crucial role in the stage of retention of a separate particle. The constant of the molecular interaction between a spherical particle and the flat surface has been estimated for the chosen value of the gap between the contacting bodies, being dependent on the wetting angle. The sufficient condition for d{sub p}-diameter particle capture by the adhesion force field (with a gap of H {approx_equal} 30 nm) is that it be brought by the appropriate

  20. Computational Fluid Dynamic Analysis of Alumina Nanofluid Coolant for a Typical PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Nazififard, Mohammad; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of); Nematollahi, Mohammad Reza [Shiraz University, Shiraz (Iran, Islamic Republic of)

    2012-05-15

    An innovative newly way of enhancing the heat transfer capability of fluids is to suspend nano-size particles in the fluid which improve the thermal conductivity of fluid. Nanofluid is a suspension of nanoparticles in base fluid. Nanofluids have attracted enormous interest from researchers due to their potential for high rate of heat exchange incurring either little or no penalty in pressure drop. Surveys such as that conducted by Williams et al. have shown that that circulation of water-based nanofluid in the primary cooling loop of PWR will improve the heat removal from the core. However, using nanofluids as working fluids has a number of limitations because any change in the reactor core materials affects the criticality and hence the effective multiplication factor. Previous studies of the application of nanofluids to LWR predicted that among nanofluids at low volume concentrations, both the alumina and zirconia nanoparticles are basically transparent to neutrons, and their contribution to coolant activation is minimal and can be used in LWRs. The scope of the present paper is to add a further contribution to nanofluids turbulent convection in a subchannel of a typical Small Modular Reactor (SMR) core. Developing turbulent forced convection flow of Al{sub 2}O{sub 3}/Water nanofluid in a subchannel is numerically investigated. The finite volume method is employed to solve the problem and two phase mixture model is considered. A three dimensional steady state is considered, with uniform heat flux on the fuel rods wall. The study is carried out for water with spherical alumina nanoparticles with a diameter of 38 nm

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  2. NONUNIFORMITIES OF TWO-PHASE COOLANT DISTRIBUTION IN A HEAT GENERATING PARTICLES BED

    Directory of Open Access Journals (Sweden)

    V. V. Sorokin

    2014-01-01

    Full Text Available Sufficient atomic power generation safety increase may be done with microfuel adapting to reactor plants with water coolant. Microfuel particle is a millimeter size grain containing fission material core in a protecting coverage. The coverage protects fuel contact with coolant and provides isolation of fission products inside. Well thermophysical properties of microfuel bed in a direct contact with water coolant excludes fuel overheating when accidents. Microfuel use was suggested for a VVER, а direct flow reactor for superheat steam generation, a reactor with neutron spectra adjustment by the steam partial content varying in the coolant.Nonuniformities of two-phase coolant distribution in a heat generating particles bed are predicted by calculations in this text. The one is due to multiple-valuedness of pressure drop across the bed on the steam quality dependency. The nonuniformity decreases with flow rate and particle size growths absolute pressure diminishing while porosity effect is weak. The worse case is for pressure quality of order of one. Some pure steam filled pores appears parallel to steam water mixture filled pores, latter steam quality is less than the mean of the bed. Considering this regime for the direct flow reactor for superheat steam generation we predict some water drops at the exit flow. The two-phase coolant filtration with subcooled water feed is unstable to strong disturbance effects are found. Uniformity of two-phase coolant distribution is worse than for one-phase in the same radial type reactor.

  3. Purification of liquid metal systems with sodium coolant from oxygen using getters

    Science.gov (United States)

    Kozlov, F. A.; Konovalov, M. A.; Sorokin, A. P.

    2016-05-01

    For increasing the safety and economic parameters of nuclear power stations (NPSs) with sodium coolant, it was decided to install all systems contacting radioactive sodium, including purification systems of circuit I, in the reactor vessel. The performance and capacity of cold traps (CTs) (conventional element of coolant purification systems) in these conditions are limited by their volume. It was proposed to use hot traps (HTs) in circuit I for coolant purification from oxygen. It was demonstrated that, at rated parameters of the installation when the temperature of the coolant streamlining the getter (gas absorber) is equal to 550°C, the hot trap can provide the required coolant purity. In shutdown modes at 250-300°C, the performance of the hot trap is reduced by four orders of magnitude. Possible HT operation regimes for shutdown modes and while reaching rated parameters were proposed and analyzed. Basic attention was paid to purification modes at power rise after commissioning and accidental contamination of the coolant when the initial oxygen concentration in it reached 25 mln-1. It was demonstrated that the efficiency of purification systems can be increased using HTs with the getter in the form of a foil or granules. The possibility of implementing the "fast purification" mode in which the coolant is purified simultaneously with passing over from the shutdown mode to the rated parameters was substantiated.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  5. Computation of Space Shuttle high-pressure cryogenic turbopump ball bearing two-phase coolant flow

    Science.gov (United States)

    Chen, Yen-Sen

    1990-01-01

    A homogeneous two-phase fluid flow model, implemented in a three-dimensional Navier-Stokes solver using computational fluid dynamics methodology is described. The application of the model to the analysis of the pump-end bearing coolant flow of the high-pressure oxygen turbopump of the Space Shuttle main engine is studied. Results indicate large boiling zones and hot spots near the ball/race contact points. The extent of the phase change of the liquid oxygen coolant flow due to the frictional and viscous heat fluxes near the contact areas has been investigated for the given inlet conditions of the coolant.

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

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse

    2012-01-01

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

  7. Behavior of primary coolant pump shaft seals during station blackout conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.C.; Rhodes, D.B.

    1986-09-12

    An assessment is made of the ability of typical Reactor Coolant Pump (RCP) Shaft Seals to withstand the conditions predicted for a station blackout (loss of all alternating current power) at a nuclear power station. Several factors are identified that are key to seal stability including inlet fluid conditions, pressure downstream of the seal, and geometrical details of the seal rings. Limits for stable seal operation are determined for various combinations of these factors, and the conclusion is drawn that some RPC seals would be near the threshold of instability during a station blackout. If the threshold were exceeded, significant leakage of coolant from the primary coolant system could be expected.

  8. Parameters important to reactor coolant pump seal stability during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.C.; Rhodes, D.B.

    1986-10-24

    An assessment is made of the ability of typical Reactor Coolant Pump (RCP) Shaft Seals to withstand the conditions predicted for a station blackout (loss of all alternating current power) at a nuclear power station. Several factors are identified that are key to seal stability including inlet fluid conditions, pressure downstream of the seal, and geometrical details of the seal rings. Limits for stable seal operation are determined for various combinations of these factors, and the conclusion is drawn that some RPC seals would be near or over the threshold of instability during a station blackout. If the threshold were exceeded, significant leakage of coolant from the primary coolant system could be expected.

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

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2015-07-01

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

  10. Single-beam thermal lens measurement of thermal diffusivity of engine coolants

    Science.gov (United States)

    George, Nibu A.; Thomas, Nibu B.; Chacko, Kavya; T, Neethu V.; Hussain Moidu, Haroon; Piyush, K.; David, Nitheesh M.

    2015-04-01

    Automobile engine coolant liquids are commonly used for efficient heat transfer from the engine to the surroundings. In this work we have investigated the thermal diffusivity of various commonly available engine coolants in Indian automobile market. We have used single beam laser induced thermal lens technique for the measurements. Engine coolants are generally available in concentrated solution form and are recommended to use at specified dilution. We have investigated the samples in the entire recommended concentration range for the use in radiators. While some of the brands show an enhanced thermal diffusivity compared to pure water, others show slight decrease in thermal diffusivity.

  11. CURRENT STATUS OF INSTRUMENTATION FOR A FLUORIDE SALT HEAT TRANSPORT DEMONSTRATION LOOP

    Energy Technology Data Exchange (ETDEWEB)

    Kisner, Roger A [ORNL; Holcomb, David Eugene [ORNL

    2010-01-01

    A small forced convection liquid fluoride salt loop is under construction at Oak Ridge National Laboratory (ORNL) to examine the heat transfer behavior of FLiNaK in a heated pebble bed. Loop operation serves several purposes: (1) reestablishing the infrastructure necessary for fluoride salt loop testing, (2) demonstrating a wireless heating technique for simulating pebble type fuel, (3) demonstration of the integration of silicon carbide (SiC) and metallic components into a liquid salt loop, and (4) demonstration of the functionality of distinctive instrumentation required for liquid fluoride salts. Loop operation requires measurement of a broad set of process variables including temperature, flow, pressure, and level. Coolant chemistry measurements (as a corrosion indicator) and component health monitoring are also important for longer-term operation. Two dominating factors in sensor and instrument selection are the high operating temperature of the salt and its chemical environment.

  12. The finite Bruck Loops

    CERN Document Server

    Baumeister, Barbara

    2009-01-01

    We continue the work by Aschbacher, Kinyon and Phillips [AKP] as well as of Glauberman [Glaub1,2] by describing the structure of the finite Bruck loops. We show essentially that a finite Bruck loop $X$ is the direct product of a Bruck loop of odd order with either a soluble Bruck loop of 2-power order or a product of loops related to the groups $PSL_2(q)$, $q= 9$ or $q \\geq 5$ a Fermat prime. The latter possibillity does occur as is shown in [Nag1, BS]. As corollaries we obtain versions of Sylow's, Lagrange's and Hall's Theorems for loops.

  13. FILM-30: A Heat Transfer Properties Code for Water Coolant

    Energy Technology Data Exchange (ETDEWEB)

    MARSHALL, THERON D.

    2001-02-01

    A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function of temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating.

  14. Study on diesel cylinder-head cooling using nanofluid coolant with jet impingement

    Directory of Open Access Journals (Sweden)

    Su Zhong-Gen

    2015-01-01

    Full Text Available To improve the heat-transfer performance of a diesel-engine cylinder head, nanofluid coolant as a new fluid was investigated, and jet impingement technology was then used to study on how to better improve heat-transfer coefficient at the nose bridge area in the diesel-engine cylinder head. Computational fluid dynamic simulation and experiments results demonstrated that using the same jet impingement parameters, the different volume shares of nanofluids showed better cooling effect than traditional coolant, but the good effect of the new cooling method was unsuitable for high volume share of nanofluid. At the same volume share of nanofluid, different jet impingement parameters such as jet angles showed different heat-transfer performance. This result implies that a strong association exists between jet impingement parameters and heat-transfer coefficient. The increase in coolant viscosity of the nanofluid coolant using jet impingement requires the expense of more drive-power cost.

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

    Science.gov (United States)

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

    2014-12-16

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

  16. Heat transfer and fluid flow aspects of fuel--coolant interactions. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M L

    1978-09-01

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

  17. Steam as coolant and lubricant in turning of metal matrix composites

    Institute of Scientific and Technical Information of China (English)

    Raviraj SHETTY; Raghuvir PAI; Vasanth KAMATH; Shrikanth S.RAO

    2008-01-01

    Green cutting has become focus of attention in ecological and environmental protection.Steam is cheap.pollution-free and eco-friendly,and then is a good and economical coolant and lubricant.Steam generator and steam feeding system were developed to generate and feed steam.Comparative experiments were carried out in cutting AA6061-15 v0l.%SiC(25 μm particle size),with cubic boron nitride(CBN)insert KB-90 grade under the conditions of compressed air,oil water emulsion,steam as coolant and lubricant,and dry cutting,respectively.The experimental results show that,with steam as coolant and lubricant,gradual reduction in the cutting force,friction coefficient,surface roughness and cutting temperature values were observed.Further,there was reduction in built up edge formation.1t is proved that use of water steam as coolant and lubricant is environmentally friendly.

  18. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  19. Turbulent Dispersion of Film Coolant and Hot Streaks in a Turbine Vane Cascade

    Science.gov (United States)

    2015-01-18

    configuration due to the large amounts of turning in the test section geometry and measurement techniques such as hot wire anemometry or temperature probe...Approved for Public Release; Distribution Unlimited Final Report: Turbulent Dispersion of Film Coolant and Hot Streaks in a Turbine Vane Cascade The...reviewed journals: Final Report: Turbulent Dispersion of Film Coolant and Hot Streaks in a Turbine Vane Cascade Report Title Magnetic resonance

  20. Power Module Cooling for Future Electric Vehicle Applications: A Coolant Comparison of Oil and PGW

    Science.gov (United States)

    2006-11-01

    POWER MODULE COOLING FOR FUTURE ELECTRIC VEHICLE APPLICATIONS: A COOLANT COMPARISON OF OIL AND PGW T. E. Salem U. S. Naval Academy 105...and efficient power converters are being developed to support the needs of future ground vehicle systems. This progress is being driven by...2006 2. REPORT TYPE N/A 3. DATES COVERED - 4. TITLE AND SUBTITLE Power Module Cooling For Future Electric Vehicle Applications: A Coolant

  1. Pseudonoise code tracking loop

    Science.gov (United States)

    Laflame, D. T. (Inventor)

    1980-01-01

    A delay-locked loop is presented for tracking a pseudonoise (PN) reference code in an incoming communication signal. The loop is less sensitive to gain imbalances, which can otherwise introduce timing errors in the PN reference code formed by the loop.

  2. Pressure drop and blower performance tests in very high temperature Helium Experimental LooP (HELP)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chan Soo; Hong, Sung Deok; Kim, Yong Wan [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Korea Atomic Energy Research Institute (KAERI) has developed the gas loops to develop and verify the key components of the nuclear hydrogen production system. At the present, KAERI is operating a small scale gas loop for feasibility tests of process heat exchanger and a very high temperature Helium Experimental LooP (HELP) for verification tests of bench scale prototypes for high temperature key components in Very High Temperature gas cooled Reactor (VHTR). Figure 1 presents the HELP assembled with the key components. The size was designed for the verification test of a 150kW intermediate heat exchanger or the simulation test in a 1/6 scaled down fuel block. The loop consists of the primary loop and the secondary loop. The primary loop and the secondary loop simulate VHTR and intermediate loop in nuclear hydrogen production system, respectively. The loops were designed to withstand the maximum temperature of 1000 .Deg. C, the maximum pressure of 9.0 MPa, and the normal mass velocity of 0.5 kg/sec. The working fluid is helium as the actual coolant of VHTR. The primary loop is composed of a preheater, a high temperature heater, a hot gas duct, intermediate heat exchangers, a water cooled U tube heat exchanger, a gas bearing circulator, a passive venting system and gas filters. The secondary loop has the same system configuration as the primary loop except a high temperature heater. Two loops share a helium supply system, a helium purification system and the water loop for a cooling tower as Figure 2. In this study, the experimental results of the bypass line pressure drop and blower performance at the nitrogen condition are analyzed to predict the main line mass flow rates without heaters.

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

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

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

  4. Experimental Study on the Effect of Late-Phase Coolant Injection on the Metallic Layer

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik; Hong, Seong Wan; Kim, Hee Dong

    2007-04-15

    Sustained heating experiments, named ELIAS (Experiments on Late-phase coolant Injection to ASsess the mitigation of focusing effect of metallic layer), were performed to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. Heat fluxes from the melt layer to the water pool varied from 250 to 550kW/m2 depending on the experimental conditions. Comparison of boiling heat fluxes between the ELIAS experiments and the calculation using the Berenson's film boiling correlation shows that effective heat removal was accomplished via late-phase coolant injection in the ELIAS experiments. In this study, simple model was developed to evaluate the mitigation of focusing effect in the metallic layer via late-phase coolant injection. The ELIAS experimental data on the heat transfer rate at the upper surface of the metallic layer were used as input data in the simple model. The calculation results for the large break loss of coolant accident in the APR1400 show that the risk induced by the focusing effect is highly dependent on the metallic layer thickness and the integrity of the reactor pressure vessel can be enhanced via late-phase coolant injection.

  5. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

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

  6. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-23

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

  8. Fuel-Coolant Interaction visualization in TROI test facility

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong-Ho; Song, Jin Ho; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    It is necessary to observe the FCI (Fuel-Coolant Interaction) phenomena at the condition of vessel failure to IVR. We carried out a visualization test on the interaction of a corium melt and water to observe the premixing phase without a free fall of a melt jet in a gas phase before contacting the cooling water. This paper is based on the previous study presented at Ninth Korea-Japan Symposium on Nuclear Hydraulics and Safety, we added the results on sieved debris distribution. The visualization test on the FCI without a free fall of a corium melt jet in a gas phase was conducted carefully in the TROI test facility. A prototypic corium consisting of uranium oxide and zirconium oxide with a weight ratio of UO{sub 2} to ZrO{sub 2} of 80 to 20, respectively, was heated up using the induction heating method. It was observed that a corium melt jet penetrated into water with 1000 mm in depth, and it took about 0.6 seconds from opening the releasing valve, which was confirmed by the sequential variation of the temperature measured by the sacrificial thermocouples installed in the direction of a falling melt jet. The cumulative mass fraction of the debris smaller than 1.0 mm was 15%, and the mass mean diameter of the debris was 2.9 mm. This visualization test can generate the valuable information such as the behavior of the corium melt jet and the size of mixing zone for validating the computer code.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-31

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

  10. Elevated-pressure mixed-coolants Joule Thomson cryocooling

    Science.gov (United States)

    Maytal, B.-Z.; Nellis, G. F.; Klein, S. A.; Pfotenhauer, J. M.

    2006-01-01

    This paper explores the potential of mixed coolants at elevated pressures for Joule-Thomson cryocooling. A numerical model of a Joule-Thomson cryocooler is developed that is capable of simulating operation with mixtures of up to 9 components consisting of hydrocarbons, non-flammable halogenated refrigerants, and inert gases. The numerical model is integrated with a genetic optimization algorithm, which has a high capability for convergence in an environment of discontinuities, constraints and local optima. The genetic optimization algorithm is used to select the optimal mixture compositions that separately maximizes following two objective functions at each elevated pressure for 80, 90 and 95 K cryocooling: the molar specific cooling capacity (the highest attainable is 3200 J/mol) and the produced cooling capacity per thermal conductance which is a measure of the compactness of the recuperator. The optimized cooling capacity for a non-flammable halogenated refrigerant mixture is smaller than for a hydrocarbon mixture; however, the cooling capacity of the two types of mixtures approach one another as pressure becomes higher. The coefficient of performance, the required heat transfer area and the effect of the number of components in the mixture is investigated as a function of the pressure. It is shown that mixtures with more components provide a higher cooling capacity but require larger recuperative heat exchangers. Optimized mixtures for 90 K cryocooling have similar cooling capacity as those for 80 K. Optimized compactness for 80 K is about 50% higher than can be achieved by pure nitrogen. For 90 K, no mixture provides a more compact recuperator than can be achieved using pure argon. The results are discussed in the context of potential applications for closed and open cycle cryocoolers.

  11. Correct numerical simulation of a two-phase coolant

    Science.gov (United States)

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

    2016-02-01

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

  12. Mass Transfer of Corrosion Products in the Nonisothermal Sodium Loop of a Fast Reactor

    Science.gov (United States)

    Varseev, E. V.; Alekseev, V. V.

    2014-11-01

    The mass transfer of the products of corrosion of the steel surface of the sodium loop of a fast nuclear power reactor was investigated for the purpose of optimization of its parameters. The problem of deposition of the corrosion products on the surface of the heat-exchange unit of the indicated loop was considered. Experimental data on the rate of accumulation of deposits in the channel of this unit and results of the dispersion analysis of the suspensions contained in the sodium coolant are presented.

  13. Hydraulic analysis of the Wendelstein 7-X cooling loops

    Energy Technology Data Exchange (ETDEWEB)

    Smirnow, M., E-mail: michael.smirnow@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Orozco, G.; Boscary, J. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Peacock, A. [European Commission c/o Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany)

    2013-10-15

    Highlights: • A hydraulic simulation model of the W7-X cooling loops and plasma facing components. • CFD analysis of orifice components. • Optimization and flow balancing of cooling loops. -- Abstract: Actively water cooled in vessel components (IVC) are required for the long pulse operation of the stellarator Wendelstein 7-X (W7-X). In total, the cooling pipes have a length of about 4.5 km, supplying the coolant via 304 cooling circuits for the IVC. Within each cooling loop, the IVC are organized mostly in parallel. A homogeneous flow through all branches or at least the minimum specified flow in all of the branches of a circuit is crucial for the IVC to withstand the loading conditions. A detailed hydraulic simulation model of the W7-X cooling loops was built with the commercial code Flowmaster, which is a 1-D computational fluid dynamics software. In order to handle the huge amount of pipe-work data that had to be modelled, a pre- and post-processing macro was developed to transfer the 3D Catia V5 CAD model to the 1-D piping model. Within this model, the hydraulic characteristics of different types of first wall components were simulated, and compared with their pressure drop measurements. As a result of this work, the need for optimization of some cooling loops has been identified and feasible modified solutions were selected.

  14. Design factors analyses of second-loop PRHRS

    Directory of Open Access Journals (Sweden)

    ZHANG Hongyan

    2017-05-01

    Full Text Available In order to study the operating characteristics of a second-loop Passive Residual Heat Removal System (PRHRS, the transient thermal analysis code RELAP5 is used to build simulation models of the main coolant system and second-loop PRHRS. Transient calculations and comparative analyses under station blackout accident and one-side feed water line break accident conditions are conducted for three critical design factors of the second-loop PRHRS:design capacity, emergency makeup tank and isolation valve opening speed. The impacts of the discussed design factors on the operating characteristics of the second-loop PRHRS are summarized based on calculations and analyses. The analysis results indicate that the system safety and cooling rate should be taken into consideration in designing PRHRS's capacity,and water injection from emergency makeup tank to steam generator can provide advantage to system cooling in the event of accident,and system startup performance can be improved by reducing the opening speed of isolation valve. The results can provide references for the design of the second-loop PRHRS in nuclear power plants.

  15. Supersymmetric Wilson loops at two loops

    CERN Document Server

    Bassetto, Antonio; Pucci, Fabrizio; Seminara, Domenico

    2008-01-01

    We study the quantum properties of certain BPS Wilson loops in ${\\cal N}=4$ supersymmetric Yang-Mills theory. They belong to a general family, introduced recently, in which the addition of particular scalar couplings endows generic loops on $S^3$ with a fraction of supersymmetry. When restricted to $S^2$, their quantum average has been further conjectured to be exactly computed by the matrix model governing the zero-instanton sector of YM$_2$ on the sphere. We perform a complete two-loop analysis on a class of cusped Wilson loops lying on a two-dimensional sphere, finding perfect agreement with the conjecture. The perturbative computation reproduces the matrix-model expectation through a highly non-trivial interplay between ladder diagrams and self-energies/vertex contributions, suggesting the existence of a localization procedure.

  16. Experimental determination of coolant flow pattern in hot and cold pools of PFBR using a large scale model

    Energy Technology Data Exchange (ETDEWEB)

    Indranil Banerjee; Rajesh, K.; AnandaRaj, M.; Venkata Ramanan, J.; Gopal, C.A.; Padmakumar, G.; Prakash, V.; Vaidyanathan, G. [Indira Gandhi Center for Atomic Research, Kalpakkam, 603102 (India)

    2005-07-01

    Full text of publication follows: The construction of Prototype Fast Breeder Reactor (PFBR) to generate 500 MWe has commenced at Kalpakkam, India. PFBR is a liquid sodium cooled pool type reactor with two secondary loops. The primary sodium pool is divided into hot pool and cold pool by means of Inner vessel. Cold sodium at 670 K is pumped through the core subassemblies and after absorbing the fission heat in the core, the sodium comes out and mixes with the hot pool at 820 K. This hot sodium exchanges heat with secondary sodium in Intermediate Heat Exchangers (IHX) which in turn transfers the heat to water in the steam generator leading to production of superheated steam to generate power. All the components like Control Plug (CP), IHX, Decay Heat Exchangers (DHX), Pump etc., are immersed in the primary sodium pool. The presence of these components influence the flow and velocity patterns of the coolant, in the hot and cold pools. The coolant behaviour in the pool is an indicator of the temperature pattern in the pool and the mechanical and thermal stresses induced on the immersed structures during transients is of significance for the safe operation of the reactor, designed for a life span of 40 years. Hence it is essential to understand the pattern of coolant flow and velocity patterns in hot and cold pools, particularly near IHX and Control plug. A 1:4 scale down model in stainless steel is constructed, simulating all the internal structures of the PFBR primary circuit for investigating the various parameters experimentally in water, to enhance the confidence in design of the primary system. The velocity distribution in the hot pool and cold pool at different regions, around the control plug, around the IHX inlet window were studied experimentally. As the coolant flow path is mainly influenced by the gravity force and inertia force, the study is conducted using Froude similitude. The magnitude of the velocity of the fluid at different points on the selected

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

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

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

  18. Study of Compatibility of Stainless Steel Weld Joints with Liquid Sodium-Potassium Coolants for Fission Surface Power Reactors for Lunar and Space Applications

    Energy Technology Data Exchange (ETDEWEB)

    Grossbeck, Martin [Univ. of Tennessee, Knoxville, TN (United States); Qualls, Louis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-07-31

    To make a manned mission to the surface of the moon or to Mars with any significant residence time, the power requirements will make a nuclear reactor the most feasible source of energy. To prepare for such a mission, NASA has teamed with the DOE to develop Fission Surface Power technology with the goal of developing viable options. The Fission Surface Power System (FSPS) recommended as the initial baseline design includes a liquid metal reactor and primary coolant system that transfers heat to two intermediate liquid metal heat transfer loops. Each intermediate loop transfers heat to two Stirling heat exchangers that each power two Stirling converters. Both the primary and the intermediate loops will use sodium-potassium (NaK) as the liquid metal coolant, and the primary loop will operate at temperatures exceeding 600°C. The alloy selected for the heat exchangers and piping is AISI Type 316L stainless steel. The extensive experience with NaK in breeder reactor programs and with earlier space reactors for unmanned missions lends considerable confidence in using NaK as a coolant in contact with stainless steel alloys. However, the microstructure, chemical segregation, and stress state of a weld leads to the potential for corrosion and cracking. Such failures have been experienced in NaK systems that have operated for times less than the eight year goal for the FSPS. For this reason, it was necessary to evaluate candidate weld techniques and expose welds to high-temperature, flowing NaK in a closed, closely controlled system. The goal of this project was to determine the optimum weld configuration for a NaK system that will withstand service for eight years under FSPS conditions. Since the most difficult weld to make and to evaluate is the tube to tube sheet weld in the intermediate heat exchangers, it was the focus of this research. A pumped loop of flowing NaK was fabricated for exposure of candidate weld specimens at temperatures of 600°C, the expected

  19. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Directory of Open Access Journals (Sweden)

    Seon Oh Yu

    2017-08-01

    Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  20. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  1. Cosmic string loop shapes

    CERN Document Server

    Blanco-Pillado, Jose J; Shlaer, Benjamin

    2015-01-01

    We analyze the shapes of cosmic string loops found in large-scale simulations of an expanding-universe string network. The simulation does not include gravitational back reaction, but we model that process by smoothing the loop using Lorentzian convolution. We find that loops at formation consist of generally straight segments separated by kinks. We do not see cusps or any cusp-like structure at the scale of the entire loop, although we do see very small regions of string that move with large Lorentz boosts. However, smoothing of the string almost always introduces two cusps on each loop. The smoothing process does not lead to any significant fragmentation of loops that were in non-self-intersecting trajectories before smoothing.

  2. Coxeter-Chein Loops

    CERN Document Server

    Blok, Rieuwert J

    2011-01-01

    In 1974 Orin Chein discovered a new family of Moufang loops which are now called Chein loops. Such a loop can be created from any group $W$ together with $\\mathbb{Z}_2$ by a variation on a semi-direct product. We study these loops in the case where $W$ is a Coxeter group and show that it has what we call a Chein-Coxeter system, a small set of generators of order 2, together with a set of relations closely related to the Coxeter relations and Chein relations. As a result we are able to give amalgam presentations for Coxeter-Chein loops. This is to our knowledge the first such presentation for a Moufang loop.

  3. Simulating Experimental Investigation on the Safety of Nuclear Heating Reactor in Loss—of —Coolant Accidents

    Institute of Scientific and Technical Information of China (English)

    ZhanjieXu

    1996-01-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology(INET) of Tisinghuan University of CHina in 1989,Its main loop is a thermal-hydraulic system with natural circulation.This paper studies the safety of NHR under the condition of loss-of -coolant accidents(LOCAs) by means of simulant experiments.First,the Background and necessity of the experiments are presented.then the experimental system,including the thermal-hydraulic system and the data collection system,and similarity criteria are introduced.Up to now ,the discharge experiments with the residual heating power(20% rated heating power)have been carried out on the experimental system,The system prameters including circulation flow rate,system pressure,system temperature,void fraction,discharge mass and so on have been recorded and analyzed.Based on the results of the experiments,the conclusionas are shown as follos:on the whole,the reactor is safe under the condition of LOCAs,but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  4. Experimental investigation of material chemical effects on emergency core cooling pump suction filter performance after loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Woon, E-mail: parkjw@dongguk.ac.k [Dongguk University, 707 Seokjang-Dong, Gyeongju, 780-714 (Korea, Republic of); Park, Byung Gi [Soonchunhyang University, Asan, Chungnam, 336-745 (Korea, Republic of); Kim, Chang Hyun [Korea Hydro and Nuclear Power Co., Ltd. 25-1, Jang-dong, Yuseong-gu, Daejeon, 305-343 (Korea, Republic of)

    2009-12-15

    Integral tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for 30 days after a loss of coolant accident. A test rig with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber and the amounts of reactor building materials are scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of calcium-silicate, reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON{sup TM} and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

  5. Simulation of fuel dispersion in the MYRRHA-FASTEF primary coolant with CFD and SIMMER-IV

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Sophia, E-mail: sophia.buckingham@vki.ac.be [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Eboli, Marica [University of Pisa, Largo Lucio Lazzarino 2, 56122 Pisa (Italy); Moreau, Vincent [CRS4, Science and Technology Park Polaris – Piscina Manna, 09010 Pula (Italy); Van Tichelen, Katrien [SCK-CEN, Boeretang 200, 2400 Mol (Belgium)

    2015-12-15

    Highlights: • A comparison between CFD and system codes applied to long-term dispersion of fuel particles inside the MYRRHA reactor is proposed. • Important accumulations at the free-surface level are to be expected. • The risk of core blockage should not be neglected. • Numerical approach and modeling assumptions have a strong influence on the simulation results and accuracy. - Abstract: The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4. Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

  6. An approach for IC engine coolant energy recovery based on low-temperature organic Rankine cycle

    Institute of Scientific and Technical Information of China (English)

    付建勤; 刘敬平; 徐政欣; 邓帮林; 刘琦

    2015-01-01

    To promote the fuel utilization efficiency of IC engine, an approach was proposed for IC engine coolant energy recovery based on low-temperature organic Rankine cycle (ORC). The ORC system uses IC engine coolant as heat source, and it is coupled to the IC engine cooling system. After various kinds of organic working media were compared, R124 was selected as the ORC working medium. According to IC engine operating conditions and coolant energy characteristics, the major parameters of ORC system were preliminary designed. Then, the effects of various parameters on cycle performance and recovery potential of coolant energy were analyzed via cycle process calculation. The results indicate that cycle efficiency is mainly influenced by the working pressure of ORC, while the maximum working pressure is limited by IC engine coolant temperature. At the same working pressure, cycle efficiency is hardly affected by both the mass flow rate and temperature of working medium. When the bottom cycle working pressure arrives at the maximum allowable value of 1.6 MPa, the fuel utilization efficiency of IC engine could be improved by 12.1%. All these demonstrate that this low-temperature ORC is a useful energy-saving technology for IC engine.

  7. Effects of staggered blades on the hydraulic characteristics of a 1400-MW canned nuclear coolant pump

    Directory of Open Access Journals (Sweden)

    Fang-Ming Zhou

    2016-08-01

    Full Text Available A canned nuclear coolant pump is used in an advanced third-generation pressurized water reactor. Impeller is a key component of a canned nuclear coolant pump. Usually, the blade is installed between the hub and the shroud as an entire part. The blade is divided into two parts and is staggered in the circumferential direction is an approach of blade design. To understand the effects of staggered blades on a canned nuclear coolant pump, this article numerically investigated different types of staggering. The validity of the numerical simulation was confirmed by comparing the numerical and experimental results. The performance change of a canned nuclear coolant pump with staggered blades was acquired. Hydraulic performance curves, axial force curves, static pressure distributions at the impeller outlet, and static pressure pulsations were performed to investigate the performance changes caused by the staggered blades. The results show that the staggered blade has an important influence on the performance of canned nuclear coolant pumps. A staggered blade does not improve hydraulic performance but does improve the axial force and pressure pulsation. Specifically, the staggered blades can significantly reduce the pressure pulsation amplitude on the impeller pass frequency.

  8. Flatness Control Using Roll Coolant Based on Predicted Flatness Variation in Cold Rolling Mills

    Science.gov (United States)

    Dohmae, Yukihiro; Okamura, Yoshihide

    Flatness control for cold rolling mills is one of the important technologies for improving of product quality and productivity. In particular, poor flatness leads to strip tearing in the extreme case and, moreover, it significantly reduces productivity. Therefore, various flatness control system has been developed. The main actuators for flatness control are classified into two types; one is mechanical equipment such as roll bender, the other is roll coolant, which controls thermal expansion of roll. Flatness variation such as center buckle or edge wave is mainly controlled by mechanical actuator which has high response characteristics. On another front, flatness variation of local zone can be controlled by roll coolant although one's response is lower than the response of mechanical actuator. For accomplishing good flatness accuracy in cold rolling mills, it is important to improve the performance of coolant control moreover. In this paper, a new coolant control method based on flatness variation model is described. In proposed method, the state of coolant spray on or off is selected to minimize the flatness deviation by using predicted flatness variation. The effectiveness of developed system has been demonstrated by application in actual plant.

  9. Coxeter-Chein Loops

    OpenAIRE

    Blok, Rieuwert J.; Gagola III, Stephen

    2011-01-01

    In 1974 Orin Chein discovered a new family of Moufang loops which are now called Chein loops. Such a loop can be created from any group $W$ together with $\\mathbb{Z}_2$ by a variation on a semi-direct product. We study these loops in the case where $W$ is a Coxeter group and show that it has what we call a Chein-Coxeter system, a small set of generators of order 2, together with a set of relations closely related to the Coxeter relations and Chein relations. As a result we are able to give am...

  10. Observational Evidence for Loop-Loop Interaction

    Science.gov (United States)

    Guiping, W.; Guangli, H.; Yuhua, T.; Aoao, X.

    2004-01-01

    Through analysis of the data including the hard x-ray(BASTE) microwave(NoRP) and magnetogram(MDI from SOHO) as well as the images of soft x-ray(YHKOH) and EIT(SOHO) on Apr. 151998 solar flare in the active region 8203(N30W12) we found: (1) there are similar quasi period oscillation in the profile of hard x-ray flux (25-5050-100keV) and microwave flux(1GHz) with duration of 85+/-25s every peak includes two sub-peak structures; (2) in the preheat phase of the flare active magnetic field changes apparently and a s-pole spot emerges ; (3) several EIT and soft x-ray loops exist and turn into bright . All of these may suggest that loop-loop interaction indeed exist. Through reconnection the electrons may be accelerated and the hard x-ray and microwave emission take place.

  11. Measurement of Coolant in a Flat Heat Pipe Using Neutron Radiography

    Science.gov (United States)

    Mizuta, Kei; Saito, Yasushi; Goshima, Takashi; Tsutsui, Toshio

    A newly developed flat heat pipe FGHPTM (Morex Kiire Co.) was experimentally investigated by using neutron radiography. The test sample of the FGHP heat spreader was 65 × 65 × 2 mm3 composed of several etched copper plates and pure water was used as the coolant. Neutron radiography was performed at the E-2 port of the Kyoto University Research Reactor (KUR). The coolant distributions in the wick area of the FGHP and its heat transfer characteristics were measured at heating conditions. Experimental results show that the coolant distributions depend slightly on its installation posture and that the liquid thickness in the wick region remains constant with increasing heat input to the FGHP. In addition, it is found that the wick surface does not dry out even in the vertical posture at present experimental conditions.

  12. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  13. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    Energy Technology Data Exchange (ETDEWEB)

    DeVan, J.H.

    1979-01-01

    Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF--BeF/sub 2/, Pb--Li alloys, and solid ceramic compounds such as Li/sub 2/O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies.

  14. What Controls DNA Looping?

    Directory of Open Access Journals (Sweden)

    Pamela J. Perez

    2014-08-01

    Full Text Available The looping of DNA provides a means of communication between sequentially distant genomic sites that operate in tandem to express, copy, and repair the information encoded in the DNA base sequence. The short loops implicated in the expression of bacterial genes suggest that molecular factors other than the naturally stiff double helix are involved in bringing the interacting sites into close spatial proximity. New computational techniques that take direct account of the three-dimensional structures and fluctuations of protein and DNA allow us to examine the likely means of enhancing such communication. Here, we describe the application of these approaches to the looping of a 92 base-pair DNA segment between the headpieces of the tetrameric Escherichia coli Lac repressor protein. The distortions of the double helix induced by a second protein—the nonspecific nucleoid protein HU—increase the computed likelihood of looping by several orders of magnitude over that of DNA alone. Large-scale deformations of the repressor, sequence-dependent features in the DNA loop, and deformability of the DNA operators also enhance looping, although to lesser degrees. The correspondence between the predicted looping propensities and the ease of looping derived from gene-expression and single-molecule measurements lends credence to the derived structural picture.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

  16. Effects of molten material temperatures and coolant temperatures on vapor explosion

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; YUAN Minghao; HU Zhihua

    2007-01-01

    An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction(FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.

  17. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Young, Michael F.

    1999-05-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-29

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

  19. The state of the art on zinc addition effect in the nuclear reactor coolant system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, U. C.; Sung, K. W.; Kim, K. R.; Paek, S.; Maeng, W. Y

    1999-12-01

    Zinc addition to the primary coolant appears promising with regard to reducing radiation dose rate, and is being used in several plants. Zinc acts to inhibit the corrosion of stainless steel by forming a thin protective film. This oxide film, with no associated increase in cobalt concentration within the film, thereby lowers the dose rate. This report on the state of art presents an overview of the zinc addition to the reactor coolant to reduce the primary system dose rate. This report discusses the effect of zinc addition for BWRs and PWRs, the thermodynamic of zinc chemistry, and the effect of zinc addition on material corrosion. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-02

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-15

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

  2. Surface Waviness in Grinding of Thin Mould Insert Using Chilled Air as Coolant

    Institute of Scientific and Technical Information of China (English)

    Yeo; S; H; K; Ramesh

    2002-01-01

    On going trend of miniaturization in electronic rel at ed parts, which is an average of two times in every 5~7 years introduce grindin g challenges. In grinding process, the surface waviness control of thin parts is an ardent task due to its warpage, induced by the high specific grinding energy (2~10 J/mm 3). Therefore, coolant is often used to avoid thermal damage, obtai n better surface integrity and to prolong wheel life. However coolant, the incomp ressibility media introduce high forces at the gri...

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

    DEFF Research Database (Denmark)

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

    2012-01-01

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

  4. Thermostat-controlled coolant pump - a new concept for fuel saving

    Energy Technology Data Exchange (ETDEWEB)

    Etemad, S. [Volvo Car Components Corp., Gothenburg (Sweden); Anderson, A. [Volvo Truck Corp., Gothenburg (Sweden)

    1999-07-01

    A new coolant pump concept has been developed for better fuel economy. The flow returning from the radiator is fed coaxially into the pump. The by-pass flow is fed tangentially into the pump, generating a pre-swirl with the same direction of rotation as the coolant pump impeller. The relative velocity between the flow and the impeller decreases. This reduces the transferred momentum from the impeller to the fluid, reducing the power consumption. The flow split between the radiator and the by-pass channel is controlled by the ordinary thermostat. Results from analysis and measurements are presented. (author)

  5. Testing loop quantum cosmology

    Science.gov (United States)

    Wilson-Ewing, Edward

    2017-03-01

    Loop quantum cosmology predicts that quantum gravity effects resolve the big-bang singularity and replace it by a cosmic bounce. Furthermore, loop quantum cosmology can also modify the form of primordial cosmological perturbations, for example by reducing power at large scales in inflationary models or by suppressing the tensor-to-scalar ratio in the matter bounce scenario; these two effects are potential observational tests for loop quantum cosmology. In this article, I review these predictions and others, and also briefly discuss three open problems in loop quantum cosmology: its relation to loop quantum gravity, the trans-Planckian problem, and a possible transition from a Lorentzian to a Euclidean space-time around the bounce point.

  6. Modern coolant additives. Environmental friendly and light metal compatible coolant additives for modern combustion engines; Moderne Kuehlmittelzusaetze. Umwelt- und leichtmetallvertraegliche Kuehlmittelzusaetze fuer moderne Verbrennungskraftmaschinen. Abschlussbericht. Vorhaben Nr. 777

    Energy Technology Data Exchange (ETDEWEB)

    Gugau, M.; Kaiser, M.

    2004-01-31

    The authors of the contribution under consideration report on the influence of the enhanced thermal stress on the impact of environmental friendly and light metal compatible coolant additives. The application and advancement of new research methods under mechanism-oriented objective led to a validation of a new guideline to the examination of the suitability of coolant additives for the coolant of internal combustion engines. Moreover, the authors create a knowledge base, on which a purposeful development can take place from suitable formulations of inhibitor for magnesium. For aluminium with silicate containing corrosion anti-freezes a close relationship between the surface temperature and the impoverishment of silicate exists. During the excess of limit temperatures, cooling agent-specific damage features arise reproducibly. The comparison of the different methods for the investigation of cavitation showed that one cannot dispense with both methods in order to evaluate a demand of insulating cavitation and a cavitative / corrosive complex regarding to the development of a test guideline. By the comprehensive electro-chemical and cavitative investigations for the magnesium alloy AZ91hp, a broad knowledge base could be formed, on which a purposeful development and evaluation of inhibitors under the use can take place from different glycols.

  7. Use of ethanolamine for alkalization of secondary coolant. First experience at VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smiesko, I. [NPP Jaslovske Bohunice (Slovakia); Bystriansky, J. [TEDIS-KOR, Dobra (Czech Republic); Szalo, A. [NPPRI Trnava (Slovakia)

    2002-07-01

    The paper summarises preparatory work and results of six-week plant trial aimed at use of ethanolamine for alkalization of secondary coolant. Operational data in pre-test and test period are given and outage inspection results are commented. Future plans are outlined. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Mao, H.; Yang, B.W.; Han, B. [Xi' an Jiaotong Univ., Shaanxi (China). Science and Technology Center for Advanced Nuclear Fuel Research

    2016-07-15

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

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

    Directory of Open Access Journals (Sweden)

    I. Kršňák

    2002-01-01

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

  10. Modeling Film-Coolant Flow Characteristics at the Exit of Shower-Head Holes

    Science.gov (United States)

    Garg, Vijay K.; Gaugler, R. E. (Technical Monitor)

    2000-01-01

    The coolant flow characteristics at the hole exits of a film-cooled blade are derived from an earlier analysis where the hole pipes and coolant plenum were also discretized. The blade chosen is the VKI rotor with three staggered rows of shower-head holes. The present analysis applies these flow characteristics at the shower-head hole exits. A multi-block three-dimensional Navier-Stokes code with Wilcox's k-omega model is used to compute the heat transfer coefficient on the film-cooled turbine blade. A reasonably good comparison with the experimental data as well as with the more complete earlier analysis where the hole pipes and coolant plenum were also gridded is obtained. If the 1/7th power law is assumed for the coolant flow characteristics at the hole exits, considerable differences in the heat transfer coefficient on the blade surface, specially in the leading-edge region, are observed even though the span-averaged values of h (heat transfer coefficient based on T(sub o)-T(sub w)) match well with the experimental data. This calls for span-resolved experimental data near film-cooling holes on a blade for better validation of the code.

  11. Vibration signal analysis of main coolant pump flywheel based on Hilbert–Huang transform

    Directory of Open Access Journals (Sweden)

    Meiru Liu

    2015-03-01

    In this paper, we present a Hilbert–Huang transform (HHT algorithm for flywheel vibration analysis. The simulation indicated that the proposed flywheel vibration signal analysis method performs well, which means that the method can lay the foundation for the detection and diagnosis in a reactor main coolant pump.

  12. Contribution to the diagnosis of mixed friction in the bearings of a reactor coolant pump

    Energy Technology Data Exchange (ETDEWEB)

    Gaev, G.P.; Shilejko, P.G.; Kail, I.T.; Proskuryakov, K.N. (Moskovskij Ehnergeticheskij Inst. (USSR)); Hippmann, N.; Kinsky, D.; Sturm, A.; Uhlemann, S. (Ingenieurhochschule Zittau (German Democratic Republic))

    1984-10-01

    Theoretical and experimental investigations have been performed to study the vibrational behaviour of a vertical, slide-bearing, fully encapsulated reactor coolant pump at various operational conditions. Magnetical and mechanical noise is interpreted as a function of pump delivery, pressure, volume flow, and temperature, and an example of an inadmissible operational condition (mixed friction in the bearings) is diagnosed.

  13. Control of oxidizing potential of Pb and Pb-Bi coolants

    Directory of Open Access Journals (Sweden)

    Vladimir Vladimirovich Ulyanov

    2015-12-01

    Full Text Available Analytical and experimental data on formation of oxygen oxidizing potential in heavy liquid metal coolants (Pb and Pb-Bi eutectic was considered. It was revealed that oxygen could be both dissolved in these coolants and included in various thermodynamically unstable oxide compounds. In case of heavy liquid metal coolant (HLMC flowing in non-isothermal circuit, these compounds are broken down with oxygen release or formed fixing dissolved oxygen. The amount of oxygen, which is present in HLMC and exhibits its activity with temperature, could be much greater than the value detected by oxygen sensor. That is why HLMC possess internal oxygen reserves inhibiting corrosion in the circuits. Presence of thermodynamically unstable oxide phases in the above coolants, non-isoconcentration distribution of active oxygen, and impossibility to currently obtain the analytical relationship showing dissolved oxygen distribution make it necessary to use at least three oxygen sensors for studying processes of formation of HLMC oxidizing potential. These sensors should be located in the zones of max and min temperatures (tmax, tmin and in that at t=450-550°С. In order to assure the most accurate estimate it is reasonable to provide additional two or more sensors in the zone at t=450-550°С.

  14. OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)

    2013-07-01

    The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.

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

    CERN Document Server

    Baek, I C; Iverson, E B

    2002-01-01

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

  16. Lead Coolant Test Facility Systems Design, Thermal Hydraulic Analysis and Cost Estimate

    Energy Technology Data Exchange (ETDEWEB)

    Soli Khericha; Edwin Harvego; John Svoboda; Ryan Dalling

    2012-01-01

    The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed: (1) Develop and Demonstrate Feasibility of Submerged Heat Exchanger; (2) Develop and Demonstrate Open-lattice Flow in Electrically Heated Core; (3) Develop and Demonstrate Chemistry Control; (4) Demonstrate Safe Operation; and (5) Provision for Future Testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimate. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-15

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

  18. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  19. Natively unstructured loops differ from other loops.

    Directory of Open Access Journals (Sweden)

    Avner Schlessinger

    2007-07-01

    Full Text Available Natively unstructured or disordered protein regions may increase the functional complexity of an organism; they are particularly abundant in eukaryotes and often evade structure determination. Many computational methods predict unstructured regions by training on outliers in otherwise well-ordered structures. Here, we introduce an approach that uses a neural network in a very different and novel way. We hypothesize that very long contiguous segments with nonregular secondary structure (NORS regions differ significantly from regular, well-structured loops, and that a method detecting such features could predict natively unstructured regions. Training our new method, NORSnet, on predicted information rather than on experimental data yielded three major advantages: it removed the overlap between testing and training, it systematically covered entire proteomes, and it explicitly focused on one particular aspect of unstructured regions with a simple structural interpretation, namely that they are loops. Our hypothesis was correct: well-structured and unstructured loops differ so substantially that NORSnet succeeded in their distinction. Benchmarks on previously used and new experimental data of unstructured regions revealed that NORSnet performed very well. Although it was not the best single prediction method, NORSnet was sufficiently accurate to flag unstructured regions in proteins that were previously not annotated. In one application, NORSnet revealed previously undetected unstructured regions in putative targets for structural genomics and may thereby contribute to increasing structural coverage of large eukaryotic families. NORSnet found unstructured regions more often in domain boundaries than expected at random. In another application, we estimated that 50%-70% of all worm proteins observed to have more than seven protein-protein interaction partners have unstructured regions. The comparative analysis between NORSnet and DISOPRED2 suggested

  20. Natively unstructured loops differ from other loops.

    Science.gov (United States)

    Schlessinger, Avner; Liu, Jinfeng; Rost, Burkhard

    2007-07-01

    Natively unstructured or disordered protein regions may increase the functional complexity of an organism; they are particularly abundant in eukaryotes and often evade structure determination. Many computational methods predict unstructured regions by training on outliers in otherwise well-ordered structures. Here, we introduce an approach that uses a neural network in a very different and novel way. We hypothesize that very long contiguous segments with nonregular secondary structure (NORS regions) differ significantly from regular, well-structured loops, and that a method detecting such features could predict natively unstructured regions. Training our new method, NORSnet, on predicted information rather than on experimental data yielded three major advantages: it removed the overlap between testing and training, it systematically covered entire proteomes, and it explicitly focused on one particular aspect of unstructured regions with a simple structural interpretation, namely that they are loops. Our hypothesis was correct: well-structured and unstructured loops differ so substantially that NORSnet succeeded in their distinction. Benchmarks on previously used and new experimental data of unstructured regions revealed that NORSnet performed very well. Although it was not the best single prediction method, NORSnet was sufficiently accurate to flag unstructured regions in proteins that were previously not annotated. In one application, NORSnet revealed previously undetected unstructured regions in putative targets for structural genomics and may thereby contribute to increasing structural coverage of large eukaryotic families. NORSnet found unstructured regions more often in domain boundaries than expected at random. In another application, we estimated that 50%-70% of all worm proteins observed to have more than seven protein-protein interaction partners have unstructured regions. The comparative analysis between NORSnet and DISOPRED2 suggested that long

  1. Introduction to Loop Heat Pipes

    Science.gov (United States)

    Ku, Jentung

    2015-01-01

    This is the presentation file for the short course Introduction to Loop Heat Pipes, to be conducted at the 2015 Thermal Fluids and Analysis Workshop, August 3-7, 2015, Silver Spring, Maryland. This course will discuss operating principles and performance characteristics of a loop heat pipe. Topics include: 1) pressure profiles in the loop; 2) loop operating temperature; 3) operating temperature control; 4) loop startup; 4) loop shutdown; 5) loop transient behaviors; 6) sizing of loop components and determination of fluid inventory; 7) analytical modeling; 8) examples of flight applications; and 9) recent LHP developments.

  2. A loop quantum multiverse?

    CERN Document Server

    Bojowald, Martin

    2013-01-01

    Inhomogeneous space-times in loop quantum cosmology have come under better control with recent advances in effective methods. Even highly inhomogeneous situations, for which multiverse scenarios provide extreme examples, can now be considered at least qualitatively.

  3. Blind loop syndrome

    Science.gov (United States)

    ... part of the stomach) and operations for extreme obesity As a complication of inflammatory bowel disease Diseases such as diabetes or scleroderma may slow down movement in a segment of the intestine, leading to blind loop syndrome.

  4. Diffusion of Wilson Loops

    CERN Document Server

    Brzoska, A M; Negele, J W; Thies, M

    2004-01-01

    A phenomenological analysis of the distribution of Wilson loops in SU(2) Yang-Mills theory is presented in which Wilson loop distributions are described as the result of a diffusion process on the group manifold. It is shown that, in the absence of forces, diffusion implies Casimir scaling and, conversely, exact Casimir scaling implies free diffusion. Screening processes occur if diffusion takes place in a potential. The crucial distinction between screening of fundamental and adjoint loops is formulated as a symmetry property related to the center symmetry of the underlying gauge theory. The results are expressed in terms of an effective Wilson loop action and compared with various limits of SU(2) Yang-Mills theory.

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

    Science.gov (United States)

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

    2009-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

  6. From Loops to Surfaces

    CERN Document Server

    Neuberger, H

    2010-01-01

    The generating function for all antisymmetric characters of a Wilson loop matrix in SU(N) Yang Mills theory is the partition function of a fermion living on the curve describing the loop. This generalizes to fermion subsystems living on higher dimensional submanifolds, for example, surfaces. This write-up also contains some extra background, in response to some questions raised during the oral presentation.

  7. Real-Time Closed Loop Modulated Turbine Cooling

    Science.gov (United States)

    Shyam, Vikram; Culley, Dennis E.; Eldridge, Jeffrey; Jones, Scott; Woike, Mark; Cuy, Michael

    2014-01-01

    It has been noted by industry that in addition to dramatic variations of temperature over a given blade surface, blade-to-blade variations also exist despite identical design. These variations result from manufacturing variations, uneven wear and deposition over the life of the part as well as limitations in the uniformity of coolant distribution in the baseline cooling design. It is proposed to combine recent advances in optical sensing, actuation, and film cooling concepts to develop a workable active, closed-loop modulated turbine cooling system to improve by 10 to 20 the turbine thermal state over the flight mission, to improve engine life and to dramatically reduce turbine cooling air usage and aircraft fuel burn. A reduction in oxides of nitrogen (NOx) can also be achieved by using the excess coolant to improve mixing in the combustor especially for rotorcraft engines. Recent patents filed by industry and universities relate to modulating endwall cooling using valves. These schemes are complex, add weight and are limited to the endwalls. The novelty of the proposed approach is twofold 1) Fluidic diverters that have no moving parts are used to modulate cooling and can operate under a wide range of conditions and environments. 2) Real-time optical sensing to map the thermal state of the turbine has never been attempted in realistic engine conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bertolotto, D.

    2011-11-15

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

  9. TACT1- TRANSIENT THERMAL ANALYSIS OF A COOLED TURBINE BLADE OR VANE EQUIPPED WITH A COOLANT INSERT

    Science.gov (United States)

    Gaugler, R. E.

    1994-01-01

    As turbine-engine core operating conditions become more severe, designers must develop more effective means of cooling blades and vanes. In order to design reliable, cooled turbine blades, advanced transient thermal calculation techniques are required. The TACT1 computer program was developed to perform transient and steady-state heat-transfer and coolant-flow analyses for cooled blades, given the outside hot-gas boundary condition, the coolant inlet conditions, the geometry of the blade shell, and the cooling configuration. TACT1 can analyze turbine blades, or vanes, equipped with a central coolant-plenum insert from which coolant-air impinges on the inner surface of the blade shell. Coolant-side heat-transfer coefficients are calculated with the heat transfer mode at each station being user specified as either impingement with crossflow, forced convection channel flow, or forced convection over pin fins. A limited capability to handle film cooling is also available in the program. The TACT1 program solves for the blade temperature distribution using a transient energy equation for each node. The nodal energy balances are linearized, one-dimensional, heat-conduction equations which are applied at the wall-outer-surface node, at the junction of the cladding and the metal node, and at the wall-inner-surface node. At the mid-metal node a linear, three-dimensional, heat-conduction equation is used. Similarly, the coolant pressure distribution is determined by solving the set of transfer momentum equations for the one-dimensional flow between adjacent fluid nodes. In the coolant channel, energy and momentum equations for one-dimensional compressible flow, including friction and heat transfer, are used for the elemental channel length between two coolant nodes. The TACT1 program first obtains a steady-state solution using iterative calculations to obtain convergence of stable temperatures, pressures, coolant-flow split, and overall coolant mass balance. Transient

  10. Development of an annular linear induction electromagnetic pump for the na-coolant circulation of LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Reyoung; Lee, Yong Bum; Kim, Yong Kyun; Nam, Ho Yun [KAERI, Taejon (Korea, Republic of)

    1998-07-01

    The EM (ElectroMagnetic) pump operated by Lorentz force (J x B) is developed for the sodium coolant circulation of LMFBR (Liquid Metal Fast Breeder Reactors). Design and experimental characterization are carried out on the linear induction EM pump of the narrow annular channel type. The pump which obtains propulsion force resultantly by the three phase symmetric alternating input currents is analyzed by the electrical equivalent circuit method used in the analyses of the induction machines. Then, the equivalent circuit for the pump consists of equivalent variables of primary and secondary resistances and magnetizing and leakage reactances given as functions of pump geometrical and electrical variables by Laithwaithe's standard formulae. Developing pressure-flowrate relation given by pump variables is sought from the balance equation on the circuit. Developing pressure and efficiency of the pump according to the pump variables are analyzed for the pump with a flowrate of 200 l/min. It is shown that pump is mainly characterized by length of the core, diameter of the inner core and channel gap geometrically and by input frequency electrically. Optimum values of pump geometrical and operational variables are determined to maximize the developing force and overall efficiency. The pump has geometrical size of 60 cm in length, 4.27 cm in inner core diameter and electrical input of 6,428 VA and 17 Hz. Optimally designed pump is manufactured by the consideration of material and operational requirements in the chemically-active sodium environment with high temperature of 600 .deg. C. Silicon-iron steel plates with high magnetic permeability in the high temperature are stacked for generation of the high magnetic flux and alumina-dispersion-strengthened-copper bands are used as exciting coils. Each turn of coil is insulated by asbestos band to protect electrical short in the high temperature. Stainless steel which can be compatible with sodium is selected as structural

  11. Genetic Programming with Simple Loops

    Institute of Scientific and Technical Information of China (English)

    QI Yuesheng; WANG Baozhong; KANG Lishan

    1999-01-01

    A kind of loop function LoopN inGenetic Programming (GP) is proposed.Different from other forms of loopfunction, such as While-Do and Repeat-Until, LoopNtakes only oneargument as its loop body and makes its loop body simply run N times,soinfinite loops will never happen. The problem of how to avoid too manylayers ofloops in Genetic Programming is also solved. The advantage ofLoopN in GP is shown bythe computational results in solving the mowerproblem.

  12. Ground experimental investigations into an ejected spray cooling system for space closed-loop application

    Directory of Open Access Journals (Sweden)

    Zhang Hongsheng

    2016-06-01

    Full Text Available Spray cooling has proved its superior heat transfer performance in removing high heat flux for ground applications. However, the dissipation of vapor–liquid mixture from the heat surface and the closed-loop circulation of the coolant are two challenges in reduced or zero gravity space environments. In this paper, an ejected spray cooling system for space closed-loop application was proposed and the negative pressure in the ejected condenser chamber was applied to sucking the two-phase mixture from the spray chamber. Its ground experimental setup was built and experimental investigations on the smooth circle heat surface with a diameter of 5 mm were conducted with distilled water as the coolant spraying from a nozzle of 0.51 mm orifice diameter at the inlet temperatures of 69.2 °C and 78.2 °C under the conditions of heat flux ranging from 69.76 W/cm2 to 311.45 W/cm2, volume flow through the spray nozzle varying from 11.22 L/h to 15.76 L/h. Work performance of the spray nozzle and heat transfer performance of the spray cooling system were analyzed; results show that this ejected spray cooling system has a good heat transfer performance and provides valid foundation for space closed-loop application in the near future.

  13. Ground experimental investigations into an ejected spray cooling system for space closed-loop application

    Institute of Scientific and Technical Information of China (English)

    Zhang Hongsheng; Li Yunze; Wang Shengnan; Liu Yang; Zhong Mingliang

    2016-01-01

    Spray cooling has proved its superior heat transfer performance in removing high heat flux for ground applications. However, the dissipation of vapor–liquid mixture from the heat sur-face and the closed-loop circulation of the coolant are two challenges in reduced or zero gravity space environments. In this paper, an ejected spray cooling system for space closed-loop application was proposed and the negative pressure in the ejected condenser chamber was applied to sucking the two-phase mixture from the spray chamber. Its ground experimental setup was built and exper-imental investigations on the smooth circle heat surface with a diameter of 5 mm were conducted with distilled water as the coolant spraying from a nozzle of 0.51 mm orifice diameter at the inlet temperatures of 69.2 ?C and 78.2 ?C under the conditions of heat flux ranging from 69.76 W/cm2 to 311.45 W/cm2, volume flow through the spray nozzle varying from 11.22 L/h to 15.76 L/h. Work performance of the spray nozzle and heat transfer performance of the spray cooling system were analyzed;results show that this ejected spray cooling system has a good heat transfer performance and provides valid foundation for space closed-loop application in the near future.

  14. Loop electrosurgical excisional procedure.

    Science.gov (United States)

    Mayeaux, E J; Harper, M B

    1993-02-01

    Loop electrosurgical excisional procedure, or LEEP, also known as loop diathermy treatment, loop excision of the transformation zone (LETZ), and large loop excision of the transformation zone (LLETZ), is a new technique for outpatient diagnosis and treatment of dysplastic cervical lesions. This procedure produces good specimens for cytologic evaluation, carries a low risk of affecting childbearing ability, and is likely to replace cryotherapy or laser treatment for cervical neoplasias. LEEP uses low-current, high-frequency electrical generators and thin stainless steel or tungsten loops to excise either lesions or the entire transformation zone. Complication rates are comparable to cryotherapy or laser treatment methods and include bleeding, incomplete removal of the lesion, and cervical stenosis. Compared with other methods, the advantages of LEEP include: removal of abnormal tissue in a manner permitting cytologic study, low cost, ease of acquiring necessary skills, and the ability to treat lesions with fewer visits. Patient acceptance of the procedure is high. Widespread use of LEEP by family physicians can be expected.

  15. Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes. 3. Investigation on diameter ratio between the coolant pipes

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-04-01

    This report presents numerical results on thermal striping characteristics at a tee junction of LMFBR coolant pipe, carried out using a direct numerical simulation code DINUS-3. In the numerical investigations, it was considered a tee junction system consisted of a main pipe (1.33 cm{sup I.D.}) with a 90deg elbow and a branch pipe having various inner diameters, and five diameter ratio conditions between both the pipes, i.e., (D{sub main}/D{sub branch}) = 1.0, 2.0, 3.0, 5.0 and 10.0. From the numerical investigations, the following characteristics were obtained: (1) Maximum sodium temperature fluctuation amplitude in the downstream region of the tee junction were decreased with increasing of the diameter ratio (decreasing of the branch pipe diameter). One of the main reasons for this behavior was considered to be that the affects of the branch pipe jet for the main pipe flows was decreased with decreasing of the branch pipe diameter. (2) Auto-power spectral density levels were decreased by the increasing of the diameter ratio. It was indicated that coolant mixing characteristics in the downstream region were controlled by locally random turbulence processes. (3) To suppress sodium temperature fluctuations in the downstream region, it is a suitable combination for larger velocity ratio and larger diameter ratio. (author)

  16. A Model for Molten Fuel-Coolant Interaction during Melt Slumping in a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sohal, Manohar Singh; Siefken, Larry James

    1999-10-01

    This paper describes a simple fuel melt slumping model to replace the current parametric model in SCDAP/RELAP5. Specifically, a fuel-coolant interaction (FCI) model is developed to analyze the slumping molten fuel, molten fuel breakup, heat transfer to coolant, relocation of the molten droplets, size of a partially solidified particles that settle to the bottom of the lower plenum, and melt-plenum interaction, if any. Considering our objectives, the molten fuel jet breakup model, and fuel droplets Lagrangian model as included in a code TEXAS-V with Eulerian thermal hydraulics for water and steam from SCDAP/RELAP5 were used. The model was assessed with experimental data from MAGICO-2000 tests performed at University of California at Santa Barbara, and FARO Test L-08 performed at Joint Research Center, Ispra, Italy. The comparison was found satisfactory.

  17. Simulation of isothermal multi-phase fuel-coolant interaction using MPS method with GPU acceleration

    Energy Technology Data Exchange (ETDEWEB)

    Gou, W.; Zhang, S.; Zheng, Y. [Zhejiang Univ., Hangzhou (China). Center for Engineering and Scientific Computation

    2016-07-15

    The energetic fuel-coolant interaction (FCI) has been one of the primary safety concerns in nuclear power plants. Graphical processing unit (GPU) implementation of the moving particle semi-implicit (MPS) method is presented and used to simulate the fuel coolant interaction problem. The governing equations are discretized with the particle interaction model of MPS. Detailed implementation on single-GPU is introduced. The three-dimensional broken dam is simulated to verify the developed GPU acceleration MPS method. The proposed GPU acceleration algorithm and developed code are then used to simulate the FCI problem. As a summary of results, the developed GPU-MPS method showed a good agreement with the experimental observation and theoretical prediction.

  18. PIV measurements of coolant flow field in a diesel engine cylinder head

    Science.gov (United States)

    Ma, Hongwei; Zhang, Zhenyang; Xue, Cheng; Huang, Yunlong

    2015-04-01

    This paper presents experimental measurements of coolant flow field in the water jacket of a diesel engine cylinder head. The test was conducted at three different flow rates using a 2-D PIV system. Appropriate tracing particles were selected and delivery device was designed and manufactured before the test. The flow parameters, such as velocity, vorticity and turbulence, were used to analyze the flow field. The effects of vortex which was located between the intake valve and the exhaust valve were discussed. The experimental results showed an asymmetric distribution of velocity in the water jacket. This led to an asymmetric thermal distribution, which would shorten the service life of the cylinder head. The structure optimization to the water jacket of cylinder head was proposed in this paper. The experimental system, especially the 2-D PIV system, is a great help to study the coolant flow structure and analyze cooling mechanism in the diesel engine cylinder head.

  19. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  20. The effect of coolants on the performance of magnetic micro-refrigerators.

    Science.gov (United States)

    Silva, D J; Bordalo, B D; Pereira, A M; Ventura, J; Oliveira, J C R E; Araújo, J P

    2014-06-01

    Magnetic refrigeration is an alternative cooling technique with envisaged technological applications on micro- and opto-electronic devices. Here, we present a magnetic micro-refrigerator cooling device with embedded micro-channels and based on the magnetocaloric effect. We studied the influence of the coolant fluid in the refrigeration process by numerically simulating the heat transfer processes using the finite element method. This allowed us to calculate the cooling power of the device. Our results show that gallium is the most efficient coolant fluid and, when used with Gd5Si2Ge2, a maximum power of 11.2 W/mm3 at a working frequency of -5 kHz can be reached. However, for operation frequencies around 50 Hz, water is the most efficient fluid with a cooling power of 0.137 W/mm3.

  1. Loops in Twistor Space

    CERN Document Server

    Bena, I; Kosower, D A; Roiban, R; Bena, Iosif; Bern, Zvi; Kosower, David A.; Roiban, Radu

    2004-01-01

    We elucidate the one-loop twistor-space structure corresponding to momentum-space MHV diagrams. We also discuss the infrared divergences, and argue that only a limited set of MHV diagrams contain them. We show how to introduce a twistor-space regulator corresponding to dimensional regularization for the infrared-divergent diagrams. We also evaluate explicitly the `holomorphic anomaly' pointed out by Cachazo, Svrcek, and Witten, and use the result to define modified differential operators which can be used to probe the twistor-space structure of one-loop amplitudes.

  2. Closed Loop Subspace Identification

    Directory of Open Access Journals (Sweden)

    Geir W. Nilsen

    2005-07-01

    Full Text Available A new three step closed loop subspace identifications algorithm based on an already existing algorithm and the Kalman filter properties is presented. The Kalman filter contains noise free states which implies that the states and innovation are uneorre lated. The idea is that a Kalman filter found by a good subspace identification algorithm will give an output which is sufficiently uncorrelated with the noise on the output of the actual process. Using feedback from the output of the estimated Kalman filter in the closed loop system a subspace identification algorithm can be used to estimate an unbiased model.

  3. Loop Quantum Gravity

    CERN Document Server

    Chiou, Dah-Wei

    2014-01-01

    This article presents an "in-a-nutshell" yet self-contained introductory review on loop quantum gravity (LQG) -- a background-independent, nonperturbative approach to a consistent quantum theory of gravity. Instead of rigorous and systematic derivations, it aims to provide a general picture of LQG, placing emphasis on the fundamental ideas and their significance. The canonical formulation of LQG, as the central topic of the article, is presented in a logically orderly fashion with moderate details, while the spin foam theory, black hole thermodynamics, and loop quantum cosmology are covered briefly. Current directions and open issues are also summarized.

  4. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    Science.gov (United States)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  5. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Shack, W.J. [Argonne National Lab., IL (United States)

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented.

  6. Primary coolant sampling for activated corrosion product studies at Hanford N Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, D.B.

    1985-01-31

    A special system for sampling primary coolant at N Reactor during operation has been constructed and operated from 1977 to 1983. The basic criteria and design for solving the difficult problem of getting representative samples have been presented; this report details how the instrumentation was configured and sampling was done. Equipment and procedures were put together to allow one person to enter a radiation zone, check on 5 monitoring instruments, operate two batch instruments, gather five partitioned samples, record 26 pieces of information, annotate a strip chart and leave the zone in 30 minutes while expending 10 mRem of exposure. Additionally, the reduction of the samples' analysis, digitization of strip chart information and storage of all data on data management systems is maintained. As built, the system provides 0.3 to 1.0 gpm streams of coolant from upstream and downstream of a steam generator. The streams are cooled to 50 to 60/sup 0/C. The radiation environment averages 20 to 50 mR/hr to the worker. Instruments and special equipment for data gathering at the sampler include pH, conductance, dissolved oxygen, dissolved hydrogen and nitrogen, hot leg and cold leg coolant temperatures, particle sizing, turbidimetry, filtration, and continuous strip chart recording.

  7. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  8. Neutronic analysis of a high power density hybrid reactor using innovative coolants

    Indian Academy of Sciences (India)

    Senay Yalçin; Mustafa Übeylı; Adem Acir

    2005-08-01

    In this study, neutronic investigation of a deuterium–tritium (DT) driven hybrid reactor using ceramic uranium fuels, namely UC, UO2 or UN under a high neutron wall load (NWL) of 10 MW/m2 at the first wall is conducted over a period of 24 months for fissile fuel breeding for light water reactors (LWRs). New substances, namely, Flinabe or Li20Sn80 are used as coolants in the fuel zone to facilitate heat transfer out of the blanket. Natural lithium is also utilized for comparison to these two innovative coolants. Neutron transport calculations are performed on a simple experimental hybrid blanket with cylindrical geometry with the help of the SCALE 4·3 System by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and an S8-P3 approximation. The investigated blanket using Flinabe or Li20Sn80 shows better fissile fuel breeding and fuel enrichment characteristics compared to that with natural lithium which shows that these two innovative coolants can be used in hybrid reactors for higher fissile fuel breeding performance. Furthermore, using a high NWL of 10 MW/m2 at the first wall of the investigated blanket can decrease the time for fuel rods to reach the level for charging in LWRs.

  9. Assessment of fiber optic sensors for aging monitoring of industrial liquid coolants

    Science.gov (United States)

    Riziotis, Christos; El Sachat, Alexandros; Markos, Christos; Velanas, Pantelis; Meristoudi, Anastasia; Papadopoulos, Aggelos

    2015-03-01

    Lately the demand for in situ and real time monitoring of industrial assets and processes has been dramatically increased. Although numerous sensing techniques have been proposed, only a small fraction can operate efficiently under harsh industrial environments. In this work the operational properties of a proposed photonic based chemical sensing scheme, capable to monitor the ageing process and the quality characteristics of coolants and lubricants in industrial heavy machinery for metal finishing processes is presented. The full spectroscopic characterization of different coolant liquids revealed that the ageing process is connected closely to the acidity/ pH value of coolants, despite the fact that the ageing process is quite complicated, affected by a number of environmental parameters such as the temperature, humidity and development of hazardous biological content as for example fungi. Efficient and low cost optical fiber sensors based on pH sensitive thin overlayers, are proposed and employed for the ageing monitoring. Active sol-gel based materials produced with various pH indicators like cresol red, bromophenol blue and chorophenol red in tetraethylorthosilicate (TEOS), were used for the production of those thin film sensitive layers deposited on polymer's and silica's large core and highly multimoded optical fibers. The optical characteristics, sensing performance and environmental robustness of those optical sensors are presented, extracting useful conclusions towards their use in industrial applications.

  10. Inverse design of a proper number, shapes, sizes, and locations of coolant flow passages

    Science.gov (United States)

    Dulikravich, George S.

    1992-01-01

    During the past several years we have developed an inverse method that allows a thermal cooling system designer to determine proper sizes, shapes, and locations of coolant passages (holes) in, say, an internally cooled turbine blade, a scram jet strut, a rocket chamber wall, etc. Using this method the designer can enforce a desired heat flux distribution on the hot outer surface of the object, while simultaneously enforcing desired temperature distributions on the same hot outer surface as well as on the cooled interior surfaces of each of the coolant passages. This constitutes an over-specified problem which is solved by allowing the number, sizes, locations and shapes of the holes to adjust iteratively until the final internally cooled configuration satisfies the over-specified surface thermal conditions and the governing equation for the steady temperature field. The problem is solved by minimizing an error function expressing the difference between the specified and the computed hot surface heat fluxes. The temperature field analysis was performed using our highly accurate boundary integral element code with linearly varying temperature along straight surface panels. Examples of the inverse design applied to internally cooled turbine blades and scram jet struts (coated and non-coated) having circular and non-circular coolant flow passages will be shown.

  11. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  12. Heat transfer performance characteristics of hybrid nanofluids as coolant in louvered fin automotive radiator

    Science.gov (United States)

    Sahoo, Rashmi R.; Sarkar, Jahar

    2016-12-01

    Present study deals with the enhancement of convective heat transfer performance of EG brine based various hybrid nanofluids i.e. Ag, Cu, SiC, CuO and TiO2 in 0-1% volume fraction of Al2O3 nanofluid, as coolants for louvered fin automobile radiator. The effects of nanoparticles combination and operating parameters on thermo physical properties, heat transfer, effectiveness, pumping power and performance index of hybrid nanofluids have been evaluated. Comparison of studied hybrid nanofluids based on radiator size and pumping power has been made as well. Among all studied hybrid nanofluids, 1% Ag hybrid nanofluid (0.5% Ag and 0.5% Al2O3) yields highest effectiveness and heat transfer rate as well as pumping power. However, SiC + Al2O3 dispersed hybrid nanofluid yields maximum performance index and hence this can be recommended for best coolant. For the same radiator size and heat transfer rate, pumping power increases by using Ag hybrid nanofluids leading to increase in engine thermal efficiency and hence reduction in engine fuel consumption. For same coolant flow rate and heat transfer rate, the radiator size reduces and pumping power increases by using Ag hybrid nanofluids leading to reduction in radiator size, weight and cost.

  13. Use of Nitrogen Trifluoride To Purify Molten Salt Reactor Coolant and Heat Transfer Fluoride Salts

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, Randall D.; Casella, Andrew M.; McNamara, Bruce K.

    2017-05-02

    Abstract: The molten salt cooled nuclear reactor is included as one of the Generation IV reactor types. One of the challenges with the implementation of this reactor is purifying and maintaining the purity of the various molten fluoride salts that will be used as coolants. The method used for Oak Ridge National Laboratory’s molten salt experimental test reactor was to treat the coolant with a mixture of H2 and HF at 600°C. In this article we evaluate thermal NF3 treatment for purifying molten fluoride salt coolant candidates based on NF3’s 1) past use to purify fluoride salts, 2) other industrial uses, 3) commercial availability, 4) operational, chemical, and health hazards, 5) environmental effects and environmental risk management methods, 6) corrosive properties, and 7) thermodynamic potential to eliminate impurities that could arise due to exposure to water and oxygen. Our evaluation indicates that nitrogen trifluoride is a viable and safer alternative to the previous method.

  14. The high-temperature sodium coolant technology in nuclear power installations for hydrogen power engineering

    Science.gov (United States)

    Kozlov, F. A.; Sorokin, A. P.; Alekseev, V. V.; Konovalov, M. A.

    2014-05-01

    In the case of using high-temperature sodium-cooled nuclear power installations for obtaining hydrogen and for other innovative applications (gasification and fluidization of coal, deep petroleum refining, conversion of biomass into liquid fuel, in the chemical industry, metallurgy, food industry, etc.), the sources of hydrogen that enters from the reactor plant tertiary coolant circuit into its secondary coolant circuit have intensity two or three orders of magnitude higher than that of hydrogen sources at a nuclear power plant (NPP) equipped with a BN-600 reactor. Fundamentally new process solutions are proposed for such conditions. The main prerequisite for implementing them is that the hydrogen concentration in sodium coolant is a factor of 100-1000 higher than it is in modern NPPs taken in combination with removal of hydrogen from sodium by subjecting it to vacuum through membranes made of vanadium or niobium. Numerical investigations carried out using a diffusion model showed that, by varying such parameters as fuel rod cladding material, its thickness, and time of operation in developing the fuel rods for high-temperature nuclear power installations (HT NPIs) it is possible to exclude ingress of cesium into sodium through the sealed fuel rod cladding. However, if the fuel rod cladding loses its tightness, operation of the HT NPI with cesium in the sodium will be unavoidable. Under such conditions, measures must be taken for deeply purifying sodium from cesium in order to minimize the diffusion of cesium into the structural materials.

  15. Heat transfer performance characteristics of hybrid nanofluids as coolant in louvered fin automotive radiator

    Science.gov (United States)

    Sahoo, Rashmi R.; Sarkar, Jahar

    2017-06-01

    Present study deals with the enhancement of convective heat transfer performance of EG brine based various hybrid nanofluids i.e. Ag, Cu, SiC, CuO and TiO2 in 0-1% volume fraction of Al2O3 nanofluid, as coolants for louvered fin automobile radiator. The effects of nanoparticles combination and operating parameters on thermo physical properties, heat transfer, effectiveness, pumping power and performance index of hybrid nanofluids have been evaluated. Comparison of studied hybrid nanofluids based on radiator size and pumping power has been made as well. Among all studied hybrid nanofluids, 1% Ag hybrid nanofluid (0.5% Ag and 0.5% Al2O3) yields highest effectiveness and heat transfer rate as well as pumping power. However, SiC + Al2O3 dispersed hybrid nanofluid yields maximum performance index and hence this can be recommended for best coolant. For the same radiator size and heat transfer rate, pumping power increases by using Ag hybrid nanofluids leading to increase in engine thermal efficiency and hence reduction in engine fuel consumption. For same coolant flow rate and heat transfer rate, the radiator size reduces and pumping power increases by using Ag hybrid nanofluids leading to reduction in radiator size, weight and cost.

  16. Safety considerations regarding the use of propane and other liquefied gases as coolants for rapid freezing purposes.

    Science.gov (United States)

    Ryan, K P; Liddicoat, M I

    1987-09-01

    Liquid propane and similar coolants are used in the rapid freezing of biological specimens. These coolants form explosive gas mixtures with air, with a 14,000-fold increase in volume over that of the liquid. The liquefied gases have high vapour pressures and, unless they are maintained below their flashpoint, the vapour above them will reach ignitable concentrations. The flashpoint of liquid propane is -104 degrees C. Ethane has a higher vapour pressure, and vapour mixed with air above liquid ethane can be ignited at a coolant temperature of -130 degrees C. The danger is minimized if the coolant is maintained near its freezing point and under a nitrogen atmosphere, in a fume cupboard. Liquid nitrogen evaporates to a 690-fold increase in volume at room temperature. It is important to ventilate the working area, especially when cryo-sectioning in a small room, otherwise there is a possibility of asphyxiation.

  17. Solar thermoelectric cooling using closed loop heat exchangers with macro channels

    Science.gov (United States)

    Atta, Raghied M.

    2017-07-01

    In this paper we describe the design, analysis and experimental study of an advanced coolant air conditioning system which cools or warms airflow using thermoelectric (TE) devices powered by solar cells. Both faces of the TE devices are directly connected to closed-loop highly efficient channels plates with macro scale channels and liquid-to-air heat exchangers. The hot side of the system consists of a pump that moves a coolant through the hot face of the TE modules, a radiator that drives heat away into the air, and a fan that transfer the heat over the radiator by forced convection. The cold side of the system consists also of a pump that moves coolant through the cold face of the TE modules, a radiator that drives cold away into the air, and a fan that blows cold air off the radiator. The system was integrated with solar panels, tested and its thermal performance was assessed. The experimental results verify the possibility of heating or cooling air using TE modules with a relatively high coefficient of performance (COP). The system was able to cool a closed space of 30 m3 by 14 °C below ambient within 90 min. The maximum COP of the whole system was 0.72 when the TE modules were running at 11.2 Å and 12 V. This improvement in the system COP over the air cooled heat sink is due to the improvement of the system heat exchange by means of channels plates.

  18. Solar thermoelectric cooling using closed loop heat exchangers with macro channels

    Science.gov (United States)

    Atta, Raghied M.

    2017-01-01

    In this paper we describe the design, analysis and experimental study of an advanced coolant air conditioning system which cools or warms airflow using thermoelectric (TE) devices powered by solar cells. Both faces of the TE devices are directly connected to closed-loop highly efficient channels plates with macro scale channels and liquid-to-air heat exchangers. The hot side of the system consists of a pump that moves a coolant through the hot face of the TE modules, a radiator that drives heat away into the air, and a fan that transfer the heat over the radiator by forced convection. The cold side of the system consists also of a pump that moves coolant through the cold face of the TE modules, a radiator that drives cold away into the air, and a fan that blows cold air off the radiator. The system was integrated with solar panels, tested and its thermal performance was assessed. The experimental results verify the possibility of heating or cooling air using TE modules with a relatively high coefficient of performance (COP). The system was able to cool a closed space of 30 m3 by 14 °C below ambient within 90 min. The maximum COP of the whole system was 0.72 when the TE modules were running at 11.2 Å and 12 V. This improvement in the system COP over the air cooled heat sink is due to the improvement of the system heat exchange by means of channels plates.

  19. Loop Quantum Gravity

    Directory of Open Access Journals (Sweden)

    Rovelli Carlo

    1998-01-01

    Full Text Available The problem of finding the quantum theory of the gravitational field, and thus understanding what is quantum spacetime, is still open. One of the most active of the current approaches is loop quantum gravity. Loop quantum gravity is a mathematically well-defined, non-perturbative and background independent quantization of general relativity, with its conventional matter couplings. Research in loop quantum gravity today forms a vast area, ranging from mathematical foundations to physical applications. Among the most significant results obtained are: (i The computation of the physical spectra of geometrical quantities such as area and volume, which yields quantitative predictions on Planck-scale physics. (ii A derivation of the Bekenstein-Hawking black hole entropy formula. (iii An intriguing physical picture of the microstructure of quantum physical space, characterized by a polymer-like Planck scale discreteness. This discreteness emerges naturally from the quantum theory and provides a mathematically well-defined realization of Wheeler's intuition of a spacetime ``foam''. Long standing open problems within the approach (lack of a scalar product, over-completeness of the loop basis, implementation of reality conditions have been fully solved. The weak part of the approach is the treatment of the dynamics: at present there exist several proposals, which are intensely debated. Here, I provide a general overview of ideas, techniques, results and open problems of this candidate theory of quantum gravity, and a guide to the relevant literature.

  20. Reversible hysteresis loop tuning

    Science.gov (United States)

    Berger, A.; Binek, Ch.; Margulies, D. T.; Moser, A.; Fullerton, E. E.

    2006-02-01

    We utilize antiferromagnetically coupled bilayer structures to magnetically tune hysteresis loop properties. Key element of this approach is the non-overlapping switching field distribution of the two magnetic layers that make up the system: a hard magnetic CoPtCrB layer (HL) and a soft magnetic CoCr layer (SL). Both layers are coupled antiferromagnetically through an only 0.6-nm-thick Ru interlayer. The non-overlapping switching field distribution allows the measurement of magnetization reversal in the SL at low fields while keeping the magnetization state of the HL unperturbed. Applying an appropriate high field or high field sequence changes the magnetic state of the HL, which then influences the SL magnetization reversal due to the interlayer coupling. In this way, the position and shape of the SL hysteresis loop can be changed or tuned in a fully reversible and highly effective manner. Here, we study specifically how the SL hysteresis loop characteristics change as we move the HL through an entire high field hysteresis loop sequence.

  1. Feasibility analysis of the Primary Loop of Pool-Type Natural Circulating Nuclear Reactor Dedicated to Seawater Desalination

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woonho; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, the feasibility of natural circulation was evaluated for the reference plant AHR400 (Advanced Heating Reactor 400MWth). AHR400 is a pool-type desalination-dedicated nuclear reactor. As a consequence, AHR400 has low operating pressure and temperature which provides large safety margin. Removal of the reactor coolant pump from the AHR400 will enforce integrity of the reactor vessel and passive safety feature. Therefore, the study also tried to find out optimized primary loop design to achieve total natural circulation of the coolant. Natural circulation capacity of the primary loop of the desalination dedicated nuclear reactor AHR400 was evaluated. It was concluded that to remove RCP from the AHR400 and operates the reactor only by natural circulation of the coolant is impossible. Decreased core power as half make removal of RCP possible with 15m central height difference between the core and IHXs. Furthermore, validation and modification of pressure loss coefficients by small-scaled natural circulation experiment at a pool-type reactor would provide more accurate results.

  2. Effect of heat release in the coolant on the stability of a water-cooled-water-moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vdovin, S.I.; Sabaev, E.F.

    1985-10-01

    The authors use exact kinetic equations in order to estimate the effect of heat release on the coolant. The authors found that the instantaneous release of even an insignificant part of the heat in the coolant exerts a significant stabilizing effect on the stability of a boiling reactor, especially in the case of a high steam content at the core outlet, which must be taken into consideration when analyzing the dynamics of boiling reactors.

  3. Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

    OpenAIRE

    Lefèvre Grégory; Živković Ljiljana S.; Jaubertie Anne

    2012-01-01

    In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspec...

  4. Experimental simulation of low rate primary coolant leaks. For the case of vessel head penetrations affected by through wall cracking

    Energy Technology Data Exchange (ETDEWEB)

    You, D.; Feron, D. [CEA-Saclay - DEN/DPC/SCCME, 91 - Gif-sur-Yvette (France); Turluer, G. [CEA-Fontenay-aux-Roses - IPSN/DES/SAMS, 92 - Fontenay-aux-Roses (France)

    2002-07-01

    An experimental simulation of primary coolant leaks was carried out to determine how the composition of the leaking liquid would change. The experiment used the EVA experimental setup, specially designed for quantitatively investigating concentration phenomena driven by evaporation. The test showed that the final composition, obtained from a solution representative of the primary coolant at the beginning of the cycle, is highly concentrated and slightly acid. The experimental results are compared with those obtained using the MULTEQ software. (authors)

  5. Always at the correct temperature. Thermal management with electric coolant pump; Immer richtig temperiert. Thermomanagement mit elektrischer Kuehlmittelpumpe

    Energy Technology Data Exchange (ETDEWEB)

    Genster, A.; Stephan, W. [Pierburg GmbH, Neuss (Germany)

    2004-11-01

    Through the use of the electric coolant pump it has become possible for the first time to attain a cooling performance which is adapted precisely to the engine load and which is independent of engine speed. For cooling the new BMW six cylinder in-line Otto engine with an engine power rating of 190 kW, the electric coolant pump by Pierburg requires only 200 W of electrical power from the onboard electrical system. (orig.)

  6. Preliminary design of a small air loop for system analysis and validation of Cathare code

    Energy Technology Data Exchange (ETDEWEB)

    Marchand, M.; Saez, M.; Tauveron, N.; Tenchine, D.; Germain, T.; Geffraye, G.; Ruby, G.P. [CEA Grenoble (DEN/DER/SSTH), 38 (France)

    2007-07-01

    The French Atomic Energy Commission (Cea) is carrying on the design of a Small Air Loop for System Analysis (SALSA), devoted to the study of gas cooled nuclear reactors behaviour in normal and incidental/accidental operating conditions. The reduced size of the SALSA components compared to a full-scale reactor and air as gaseous coolant instead of Helium will allow an easy management of the loop. The main purpose of SALSA will be the validation of the associated thermal hydraulic safety simulation codes, like CATHARE. The main goal of this paper is to present the methodology used to define the characteristics of the loop. In a first step, the study has been focused on a direct-cycle system for the SALSA loop with few global constraints using a similarity analysis to support the definition and design of the loop. Similarity requirements have been evaluated to determine the scale factors which have to be applied to the SALSA loop components. The preliminary conceptual design of the SALSA plant with a definition of each component has then be carried out. The whole plant has been modelled using the CATHARE code. Calculations of the SALSA steady-state in nominal conditions and of different plant transients in direct-cycle have been made. The first system results obtained on the global behaviour of the loop confirm that SALSA can be representative of a Gas-Cooled nuclear reactor with some minor design modifications. In a second step, the current prospects focus on the SALSA loop capability to reproduce correctly the heat transfer occurring in specific incidental situations. Heat decay removal by natural convection is a crucial point of interest. The first results show that the behaviour and the efficiency of the loop are strongly influenced by the definition of the main parameters for each component. A complete definition of SALSA is under progress. (authors)

  7. Breakup of jet and drops during premixing phase of fuel coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Haraldsson, Haraldur Oskar

    2000-05-01

    During the course of a hypothetical severe accident in a light water reactor, molten liquid may be introduced into a volatile coolant, which, under certain conditions, results in explosive interactions. Such fuel-coolant interactions (FCI) are characterised by an initial pre-mixing phase during which the molten liquid, metallic or oxidic in nature, undergoes a breakup (fragmentation) process which significantly increase the area available for melt-coolant contact, and thus energy transfer. Although substantial progress in the understanding of phenomenology of the FCI events has been achieved in recent years, there remain uncertainties in describing the primary and secondary breakup processes. The focus of this work is on the melt jet and drop breakup during the premixing phase of FCI. The objectives are to gain insight into the premixing phase of the FCI phenomena, to determine what fraction of the melt fragments and determine the size distribution. The approach is to perform experiments with various simulant materials, at different scales, different conditions and with variation of controlling parameters affecting jet and drop breakup processes. The analysis approach is to investigate processes at different level of detail and complexity to understand the physics, to rationalise experimental results and to develop and validate models. In the first chapter a brief introduction and review of the status of the FCI phenomena is performed. A review of previous and current experimental projects is performed. The status of the experimental projects and major findings are outlined. The first part of the second chapter deals with experimental investigation of jet breakup. Two series of experiments were performed with low and high temperature jets. The low temperature experiments employed cerrobend-70 as jet liquid. A systematic investigation of thermal hydraulic conditions and melt physical properties on the jet fragmentation and particle debris characteristics was

  8. Two-loop and n-loop eikonal vertex corrections

    OpenAIRE

    Kidonakis, Nikolaos

    2003-01-01

    I present calculations of two-loop vertex corrections with massive and massless partons in the eikonal approximation. I show that the $n$-loop result for the UV poles can be given in terms of the one-loop calculation.

  9. Conceptual Design of Forced Convection Molten Salt Heat Transfer Testing Loop

    Energy Technology Data Exchange (ETDEWEB)

    Manohar S. Sohal; Piyush Sabharwall; Pattrick Calderoni; Alan K. Wertsching; S. Brandon Grover

    2010-09-01

    This report develops a proposal to design and construct a forced convection test loop. A detailed test plan will then be conducted to obtain data on heat transfer, thermodynamic, and corrosion characteristics of the molten salts and fluid-solid interaction. In particular, this report outlines an experimental research and development test plan. The most important initial requirement for heat transfer test of molten salt systems is the establishment of reference coolant materials to use in the experiments. An earlier report produced within the same project highlighted how thermophysical properties of the materials that directly impact the heat transfer behavior are strongly correlated to the composition and impurities concentration of the melt. It is therefore essential to establish laboratory techniques that can measure the melt composition, and to develop purification methods that would allow the production of large quantities of coolant with the desired purity. A companion report describes the options available to reach such objectives. In particular, that report outlines an experimental research and development test plan that would include following steps: •Molten Salts: The candidate molten salts for investigation will be selected. •Materials of Construction: Materials of construction for the test loop, heat exchangers, and fluid-solid corrosion tests in the test loop will also be selected. •Scaling Analysis: Scaling analysis to design the test loop will be performed. •Test Plan: A comprehensive test plan to include all the tests that are being planned in the short and long term time frame will be developed. •Design the Test Loop: The forced convection test loop will be designed including extensive mechanical design, instrument selection, data acquisition system, safety requirements, and related precautionary measures. •Fabricate the Test Loop. •Perform the Tests. •Uncertainty Analysis: As a part of the data collection, uncertainty analysis will

  10. Local loop near-rings

    OpenAIRE

    Franetič, Damir

    2015-01-01

    We study loop near-rings, a generalization of near-rings, where the additive structure is not necessarily associative. We introduce local loop near-rings and prove a useful detection principle for localness.

  11. Boosted Fast Flux Loop Alternative Cooling Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Glen R. Longhurst; Donna Post Guillen; James R. Parry; Douglas L. Porter; Bruce W. Wallace

    2007-08-01

    The Gas Test Loop (GTL) Project was instituted to develop the means for conducting fast neutron irradiation tests in a domestic radiation facility. It made use of booster fuel to achieve the high neutron flux, a hafnium thermal neutron absorber to attain the high fast-to-thermal flux ratio, a mixed gas temperature control system for maintaining experiment temperatures, and a compressed gas cooling system to remove heat from the experiment capsules and the hafnium thermal neutron absorber. This GTL system was determined to provide a fast (E > 0.1 MeV) flux greater than 1.0E+15 n/cm2-s with a fast-to-thermal flux ratio in the vicinity of 40. However, the estimated system acquisition cost from earlier studies was deemed to be high. That cost was strongly influenced by the compressed gas cooling system for experiment heat removal. Designers were challenged to find a less expensive way to achieve the required cooling. This report documents the results of the investigation leading to an alternatively cooled configuration, referred to now as the Boosted Fast Flux Loop (BFFL). This configuration relies on a composite material comprised of hafnium aluminide (Al3Hf) in an aluminum matrix to transfer heat from the experiment to pressurized water cooling channels while at the same time providing absorption of thermal neutrons. Investigations into the performance this configuration might achieve showed that it should perform at least as well as its gas-cooled predecessor. Physics calculations indicated that the fast neutron flux averaged over the central 40 cm (16 inches) relative to ATR core mid-plane in irradiation spaces would be about 1.04E+15 n/cm2-s. The fast-to-thermal flux ratio would be in excess of 40. Further, the particular configuration of cooling channels was relatively unimportant compared with the total amount of water in the apparatus in determining performance. Thermal analyses conducted on a candidate configuration showed the design of the water coolant and

  12. Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J. [Brookhaven National Lab., Upton, NY (United States)

    1998-03-01

    Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs.

  13. On the extended loop calculus

    CERN Document Server

    Griego, J R

    1995-01-01

    Some features of extended loops are considered. In particular, the behaviour under diffeomorphism transformations of the wavefunctions with support on the extended loop space are studied. The basis of a method to obtain analytical expressions of diffeomorphism invariants via extended loops are settled. Applications to knot theory and quantum gravity are considered.

  14. Closing global material loops

    DEFF Research Database (Denmark)

    Prosman, Ernst-Jan; Wæhrens, Brian Vejrum; Liotta, Giacomo

    2017-01-01

    Replacing virgin materials with waste materials, a practice known as Industrial Symbiosis (IS), has been identified as a key strategy for closing material loops. This article adopts a critical view on geographic proximity and external coordinators – two key enablers of IS. By ‘uncovering’ a case...... where both enablers are absent, this study seeks to explore firm-level challenges of IS. We adopt an exploratory case study approach at a cement manufacturer who engages in cross-border IS without the support of external coordinators. Our research presents insights into two key areas of IS: 1) setting...... for geographic proximity and external coordinators. In doing so, our insights into firm-level challenges of long-distance IS exchanges contribute to closing global material loops by increasing the number of potential circular pathways....

  15. Loop Quantum Cosmology

    Directory of Open Access Journals (Sweden)

    Bojowald Martin

    2008-07-01

    Full Text Available Quantum gravity is expected to be necessary in order to understand situations in which classical general relativity breaks down. In particular in cosmology one has to deal with initial singularities, i.e., the fact that the backward evolution of a classical spacetime inevitably comes to an end after a finite amount of proper time. This presents a breakdown of the classical picture and requires an extended theory for a meaningful description. Since small length scales and high curvatures are involved, quantum effects must play a role. Not only the singularity itself but also the surrounding spacetime is then modified. One particular theory is loop quantum cosmology, an application of loop quantum gravity to homogeneous systems, which removes classical singularities. Its implications can be studied at different levels. The main effects are introduced into effective classical equations, which allow one to avoid the interpretational problems of quantum theory. They give rise to new kinds of early-universe phenomenology with applications to inflation and cyclic models. To resolve classical singularities and to understand the structure of geometry around them, the quantum description is necessary. Classical evolution is then replaced by a difference equation for a wave function, which allows an extension of quantum spacetime beyond classical singularities. One main question is how these homogeneous scenarios are related to full loop quantum gravity, which can be dealt with at the level of distributional symmetric states. Finally, the new structure of spacetime arising in loop quantum gravity and its application to cosmology sheds light on more general issues, such as the nature of time.

  16. Loop Quantum Cosmology

    Directory of Open Access Journals (Sweden)

    Bojowald Martin

    2005-12-01

    Full Text Available Quantum gravity is expected to be necessary in order to understand situations where classical general relativity breaks down. In particular in cosmology one has to deal with initial singularities, i.e., the fact that the backward evolution of a classical space-time inevitably comes to an end after a finite amount of proper time. This presents a breakdown of the classical picture and requires an extended theory for a meaningful description. Since small length scales and high curvatures are involved, quantum effects must play a role. Not only the singularity itself but also the surrounding space-time is then modified. One particular realization is loop quantum cosmology, an application of loop quantum gravity to homogeneous systems, which removes classical singularities. Its implications can be studied at different levels. Main effects are introduced into effective classical equations which allow to avoid interpretational problems of quantum theory. They give rise to new kinds of early universe phenomenology with applications to inflation and cyclic models. To resolve classical singularities and to understand the structure of geometry around them, the quantum description is necessary. Classical evolution is then replaced by a difference equation for a wave function which allows to extend space-time beyond classical singularities. One main question is how these homogeneous scenarios are related to full loop quantum gravity, which can be dealt with at the level of distributional symmetric states. Finally, the new structure of space-time arising in loop quantum gravity and its application to cosmology sheds new light on more general issues such as time.

  17. PAR Loop Schedule Review

    Energy Technology Data Exchange (ETDEWEB)

    Schaffer, Jr.; W.F.

    1958-04-30

    The schedule for the installation of the PAR slurry loop experiment in the South Facility of the ORR has been reviewed and revised. The design, fabrications and Installation is approximately two weeks behind schedule at this time due to many factors; however, indications are that this time can be made up. Design is estimated to be 75% complete, fabrication 32% complete and installation 12% complete.

  18. Verification of Loop Diagnostics

    Science.gov (United States)

    Winebarger, A.; Lionello, R.; Mok, Y.; Linker, J.; Mikic, Z.

    2014-01-01

    Many different techniques have been used to characterize the plasma in the solar corona: density-sensitive spectral line ratios are used to infer the density, the evolution of coronal structures in different passbands is used to infer the temperature evolution, and the simultaneous intensities measured in multiple passbands are used to determine the emission measure. All these analysis techniques assume that the intensity of the structures can be isolated through background subtraction. In this paper, we use simulated observations from a 3D hydrodynamic simulation of a coronal active region to verify these diagnostics. The density and temperature from the simulation are used to generate images in several passbands and spectral lines. We identify loop structures in the simulated images and calculate the loop background. We then determine the density, temperature and emission measure distribution as a function of time from the observations and compare with the true temperature and density of the loop. We find that the overall characteristics of the temperature, density, and emission measure are recovered by the analysis methods, but the details of the true temperature and density are not. For instance, the emission measure curves calculated from the simulated observations are much broader than the true emission measure distribution, though the average temperature evolution is similar. These differences are due, in part, to inadequate background subtraction, but also indicate a limitation of the analysis methods.

  19. Cosmic string loop microlensing

    Science.gov (United States)

    Bloomfield, Jolyon K.; Chernoff, David F.

    2014-06-01

    Cosmic superstring loops within the galaxy microlens background point sources lying close to the observer-string line of sight. For suitable alignments, multiple paths coexist and the (achromatic) flux enhancement is a factor of two. We explore this unique type of lensing by numerically solving for geodesics that extend from source to observer as they pass near an oscillating string. We characterize the duration of the flux doubling and the scale of the image splitting. We probe and confirm the existence of a variety of fundamental effects predicted from previous analyses of the static infinite straight string: the deficit angle, the Kaiser-Stebbins effect, and the scale of the impact parameter required to produce microlensing. Our quantitative results for dynamical loops vary by O(1) factors with respect to estimates based on infinite straight strings for a given impact parameter. A number of new features are identified in the computed microlensing solutions. Our results suggest that optical microlensing can offer a new and potentially powerful methodology for searches for superstring loop relics of the inflationary era.

  20. Transmutation performance analysis on coolant options in a hybrid reactor system design for high level waste incineration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong-Hee; Siddique, Muhammad Tariq; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2015-11-15

    Highlights: • Waste transmutation performance was compared and analyzed for seven different coolant options. • Reactions of fission and capture showed big differences depending on coolant options. • Moderation effect significantly affects on energy multiplication, tritium breeding and waste transmutation. • Reduction of radio-toxicities of TRUs showed different trend to coolant choice from performance of waste transmutation. - Abstract: A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H{sub 2}O, D{sub 2}O, Na, PbBi, LiF-BeF{sub 2} and NaF-BeF{sub 2} applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are k{sub eff}, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H{sub 2}O and D{sub 2}O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF{sub 2} and NaF-BeF{sub 2} is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

  1. LoopIng: a template-based tool for predicting the structure of protein loops.

    KAUST Repository

    Messih, Mario Abdel

    2015-08-06

    Predicting the structure of protein loops is very challenging, mainly because they are not necessarily subject to strong evolutionary pressure. This implies that, unlike the rest of the protein, standard homology modeling techniques are not very effective in modeling their structure. However, loops are often involved in protein function, hence inferring their structure is important for predicting protein structure as well as function.We describe a method, LoopIng, based on the Random Forest automated learning technique, which, given a target loop, selects a structural template for it from a database of loop candidates. Compared to the most recently available methods, LoopIng is able to achieve similar accuracy for short loops (4-10 residues) and significant enhancements for long loops (11-20 residues). The quality of the predictions is robust to errors that unavoidably affect the stem regions when these are modeled. The method returns a confidence score for the predicted template loops and has the advantage of being very fast (on average: 1 min/loop).www.biocomputing.it/loopinganna.tramontano@uniroma1.itSupplementary data are available at Bioinformatics online.

  2. Resume of KER Loop irradiation tests for 1957, 1958, and 1959

    Energy Technology Data Exchange (ETDEWEB)

    Kratzer, W.K.

    1960-03-28

    In 1957 three of the four high temperature recirculating loops in KE Reactor were put into service and in 1958 the fourth loop was started up. The primary purpose of the loops has been high temperature irradiation testing of fuel elements in support of the NPR fuel element development program, although tests have been run to provide information for PRTR fuel development and high temperature aluminum corrosion programs. Coolant technology studies relating to activity build-up, crude formation, corrosion, failure detection, and decontamination are carried on in conjunction with the fuel element tests. The purpose of this document is to provide a resume of all fuel element test irradiations in the KER Loops from the time they started up to the end of CY-1959. A description of each test, its operating conditions, a chronological history of each loop, and a bibliography of documents pertaining to the tests are included. Supplements to this report will be issued periodically to bring the information up to date.

  3. A passively-safe fusion reactor blanket with helium coolant and steel structure

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, K.M.

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  4. Proceedings of the CSNI specialists meeting on fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    None

    1994-03-01

    A specialists meeting on fuel-coolant interactions was held in Santa Barbara, CA from January 5-7, 1993. The meeting was sponsored by the United States Nuclear Regulatory Commission in collaboration with the Committee on the Safety of Nuclear Installation (CSNI) of the OECD Nuclear Energy Agency (NEA) and the University of California at Santa Barbara. The objectives of the meeting are to cross-fertilize on-going work, provide opportunities for mutual check points, seek to focus the technical issues on matters of practical significance and re-evaluate both the objectives as well as path of future research. Individual papers have been cataloged separately.

  5. Physical properties of heavy liquid-metal coolants in a wide temperature range

    Directory of Open Access Journals (Sweden)

    Borisenko A.

    2011-05-01

    Full Text Available The pulse-phase method, the gamma-attenuation method and the method of dumping oscillation of a crucible with a melt were used for measuring the velocity of sound, the density and the kinematic viscosity of a set of liquid-metal coolants for perspective nuclear reactors. There are liquid gallium, indium, tin, lead, bismuth and lead-bismuth eutectic alloy among the melts investigated. The accuracy of the measurements was as high as 0.3%, 0.2 to 0.4% and 1.5% for the ultrasound velocity, the density and the viscosity, correspondingly.

  6. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Peyrouty, P.

    1996-12-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

  7. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  8. The development of Sn-Li coolant/breeding material for APEX/ALPS applications.

    Energy Technology Data Exchange (ETDEWEB)

    Sze, D.-K.

    1999-07-08

    A Sn-Li alloy has been identified to be a coolant/breeding material for D-T fusion applications. The key feature of this material is its very low vapor pressure, which will be very useful for free surface concepts employed in APEX, ALPS and inertial confinement fission. The vapor is dominated by lithium, which has very low Z. Initial assessment of the material indicates acceptable tritium breeding capability, high thermal conductivity, expected low tritium volubility, and expected low chemical reactivities with water and air. Some key concerns are the high activation and material compatibility issues. The initial assessment of this material, for fission applications, is presented in this paper.

  9. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

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

  10. Improved solidification influence modelling for Eulerian fuel-coolant interaction codes

    Energy Technology Data Exchange (ETDEWEB)

    Ursic, Mitja, E-mail: mitja.ursic@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Leskovar, Matjaz; Mavko, Borut [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    Steam explosion experiments revealed important differences in the efficiency between simulant alumina and oxidic corium melts. The experimentally observed differences are importantly attributed to the differences in the melt droplets solidification and void production, which are limiting phenomena in the steam explosion process and have to be adequately modelled in fuel-coolant interaction codes. This article focuses on the modelling of the solidification effect. An improved solidification influence modelling approach for Eulerian fuel-coolant interaction codes was developed and is presented herein. The solidification influence modelling in fuel-coolant interaction codes is strongly related to the modelling of the temperature profile and the mechanical effect of the crust on the fragmentation process. Therefore the first objective was to introduce an improved temperature profile modelling and a fragmentation criterion for partly solidified droplets. The fragmentation criterion was based on the established modified Weber number, which considers the crust stiffness as a stabilizing force acting to retain the crust under presence of the hydrodynamic forces. The modified Weber number was validated on experimental data. The application of the developed improved solidification influence modelling enables an improved determination of the melt droplet mass, which can be efficiently involved in the fine fragmentation during the steam explosion process. Additionally, also the void production modelling is improved, because it is strongly related to the temperature profile modelling in the frame of the solidification influence modelling. Therefore the second objective was to enable an improved solidification influence modelling in codes with an Eulerian formulation of the droplet field. Two additional transported model parameters based on the most important droplets features regarding the fuel-coolant interaction behaviour, were derived. First, the crust stiffness was

  11. Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

    Energy Technology Data Exchange (ETDEWEB)

    Zapata, Luis E. (Livermore, CA); Hackel, Lloyd (Livermore, CA)

    1999-01-01

    A lamp system with a very soft high-intensity output is provided over a large area by water cooling a long-arc lamp inside a diffuse reflector of polytetrafluorethylene (PTFE) and titanium dioxide (TiO.sub.2) white pigment. The water is kept clean and pure by a one micron particulate filter and an activated charcoal/ultraviolet irradiation system that circulates and de-ionizes and biologically sterilizes the coolant water at all times, even when the long-arc lamp is off.

  12. Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

    Energy Technology Data Exchange (ETDEWEB)

    Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

    1998-01-01

    This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

  13. Estimation of the coolant flow through a natural circulation BWR fuel channel applying and equivalent electrical model

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Campus Morelos en IMTA, Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_ig@yahoo.com.m [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    This work presents the design and implementation of an advanced controller for a reduced order model of a BWR reactor core cooled by natural circulating water, which allows real time estimates of coolant flows through fuel assemblies about standard neutron flux strings. Nuclear power plants with boiling water reactors control individual fuel assembly coolant flows by forced circulation using external or internal water pumps and different core support plate orifices. These two elements reduce flow dependency on local channel pressure drops. In BWR reactors using only natural circulation coolant flows, these two elements are not available and therefore individual channel coolant flows are highly dependent in local conditions, such as power distributions and local pressure drops. Therefore it is expected that grater uncertainties in these variables be used during safety, fuel management and other analysis, which in turns may lead to increased operation penalties, such as tighter operating limits. The objective of this work is to asses by computer simulations means to reduce uncertainties in the measurement of fuel assembly coolant flows and eventually the associated penalties. During coolant phase transitions, pressure drops and local power may alter local natural circulation through fuel assemblies and flow estimates can be helped or not by control rod moves. This work presents the construction of an optimal controller for a core flow estimator based on a reduced order model of the coolant going though the reactor vessel components and nuclear core. This model is to be driven by plant signals from standard BWR instrumentation in order to estimate the coolant flows in selected fuel assemblies about a LPRM string. For this purpose an equivalent electrical model has been mathematically developed and numerically tested. The power-flow maps of typical BRW are used as steady state references for this equivalent model. Once these were fully reproduced for steady state

  14. The influence of EI-21 redox ion-exchange resins on the secondary-coolant circuit water chemistry of vehicular nuclear power installations

    Science.gov (United States)

    Moskvin, L. N.; Rakov, V. T.

    2015-06-01

    The results obtained from testing the secondary-coolant circuit water chemistry of full-scale land-based prototype bench models of vehicular nuclear power installations equipped with water-cooled water-moderated and liquid-metal reactor plants are presented. The influence of copper-containing redox ionexchange resins intended for chemically deoxygenating steam condensate on the working fluid circulation loop's water chemistry is determined. The influence of redox ion-exchange resins on the water chemistry is evaluated by generalizing an array of data obtained in the course of extended monitoring using the methods relating to physicochemical analysis of the quality of condensate-feedwater path media and the methods relating to metallographic analysis of the state of a faulty steam generator's tube system surfaces. The deoxygenating effectiveness of the normal state turbine condensate vacuum deaeration system is experimentally determined. The refusal from applying redox ion-exchange resins in the condensate polishing ion-exchange filters is formulated based on the obtained data on the adverse effect of copper-containing redox ionexchange resins on the condensate-feedwater path water chemistry and based on the data testifying a sufficient effect from using the normal state turbine condensate vacuum deaeration system. Data on long-term operation of the prototype bench model of a vehicular nuclear power installation without subjecting the turbine condensate to chemical deoxygenation are presented.

  15. The premixing and propagation phases of fuel-coolant interactions: a review of recent experimental studies and code developments

    Energy Technology Data Exchange (ETDEWEB)

    Antariksawan, A.R. [Reactor Safety Technology Research Center of BATAN (Indonesia); Moriyama, Kiyofumi; Park, Hyun-sun; Maruyama, Yu; Yang, Yanhua; Sugimoto, Jun

    1998-09-01

    A vapor explosion (or an energetic fuel-coolant interactions, FCIs) is a process in which hot liquid (fuel) transfers its internal energy to colder, more volatile liquid (coolant); thus the coolant vaporizes at high pressure and expands and does works on its surroundings. Traditionally, the energetic fuel-coolant interactions could be distinguished in subsequent stages: premixing (or coarse mixing), triggering, propagation and expansion. Realizing that better and realistic prediction of fuel-coolant interaction consequences will be available understanding the phenomenology in the premixing and propagation stages, many experimental and analytical studies have been performed during more than two decades. A lot of important achievements are obtained during the time. However, some fundamental aspects are still not clear enough; thus the works are directed to that direction. In conjunction, the model/code development is pursuit. This is aimed to provide a scaling tool to bridge the experimental results to the real geometries, e.g. reactor pressure vessel, reactor containment. The present review intends to collect the available information on the recent works performed to study the premixing and propagation phases. (author). 97 refs.

  16. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  17. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  18. Improving Loop Dependence Analysis

    DEFF Research Database (Denmark)

    Jensen, Nicklas Bo; Karlsson, Sven

    2017-01-01

    Programmers can no longer depend on new processors to have significantly improved single-thread performance. Instead, gains have to come from other sources such as the compiler and its optimization passes. Advanced passes make use of information on the dependencies related to loops. We improve...... the quality of that information by reusing the information given by the programmer for parallelization. We have implemented a prototype based on GCC into which we also add a new optimization pass. Our approach improves the amount of correctly classified dependencies resulting in 46% average improvement...

  19. Closing the loop.

    Science.gov (United States)

    Dassau, E; Atlas, E; Phillip, M

    2011-02-01

    Closed-loop algorithms can be found in every aspect of everyday modern life. Automation and control are used constantly to provide safety and to improve quality of life. Closed-loop systems and algorithms can be found in home appliances, automobiles, aviation and more. Can one imagine nowadays driving a car without ABS, cruise control or even anti-sliding control? Similar principles of automation and control can be used in the management of diabetes mellitus (DM). The idea of an algorithmic/technological way to control glycaemia is not new and has been researched for more than four decades. However, recent improvements in both glucose-sensing technology and insulin delivery together with advanced control and systems engineering made this dream of an artificial pancreas possible. The artificial pancreas may be the next big step in the treatment of DM since the use of insulin analogues. An artificial pancreas can be described as internal or external devices that use continuous glucose measurements to automatically manage exogenous insulin delivery with or without other hormones in an attempt to restore glucose regulation in individuals with DM using a control algorithm. This device as described can be internal or external; can use different types of control algorithms with bi-hormonal or uni-hormonal design; and can utilise different ways to administer them. The different designs and implementations have transitioned recently from in silico simulations to clinical evaluation stage with practical applications in mind. This may mark the beginning of a new era in diabetes management with the introduction of semi-closed-loop systems that can prevent or minimise nocturnal hypoglycaemia, to hybrid systems that will manage blood glucose (BG) levels with minimal user intervention to finally fully automated systems that will take the user out of the loop. More and more clinical trials will be needed for the artificial pancreas to become a reality but initial encouraging

  20. Method for calculating coolant resonance frequencies under normal and accident conditions in nuclear power plants with WWER-type pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N. (Moskovskij Ehnergeticheskij Inst. (USSR))

    1983-03-01

    Mathematical models are proposed for calculating acoustic oscillation resonance frequencies in the coolant in various components of the WWER type primary circuit (core, steam generator, pressurizer, piping). Due to the correspondence between model calculations and experimental results obtained in operating nuclear power plants, the developed models can be used for practical calculations. The possibility of calculating the eigenfrequencies of the coolant oscillation under different operating conditions leads to the interpretation of operational data, to the analysis of operational conditions, to the detection of coolant boiling in the reactor, and to design changes in order to prevent resonance oscillations within the coolant.

  1. Loop expansion and the bosonic representation of loop quantum gravity

    Science.gov (United States)

    Bianchi, E.; Guglielmon, J.; Hackl, L.; Yokomizo, N.

    2016-10-01

    We introduce a new loop expansion that provides a resolution of the identity in the Hilbert space of loop quantum gravity on a fixed graph. We work in the bosonic representation obtained by the canonical quantization of the spinorial formalism. The resolution of the identity gives a tool for implementing the projection of states in the full bosonic representation onto the space of solutions to the Gauss and area matching constraints of loop quantum gravity. This procedure is particularly efficient in the semiclassical regime, leading to explicit expressions for the loop expansions of coherent, heat kernel and squeezed states.

  2. Loop expansion and the bosonic representation of loop quantum gravity

    CERN Document Server

    Bianchi, Eugenio; Hackl, Lucas; Yokomizo, Nelson

    2016-01-01

    We introduce a new loop expansion that provides a resolution of the identity in the Hilbert space of loop quantum gravity on a fixed graph. We work in the bosonic representation obtained by the canonical quantization of the spinorial formalism. The resolution of the identity gives a tool for implementing the projection of states in the full bosonic representation onto the space of solutions to the Gauss and area matching constraints of loop quantum gravity. This procedure is particularly efficient in the semiclassical regime, leading to explicit expressions for the loop expansions of coherent, heat kernel and squeezed states.

  3. TRANSPORTATION MODAL CHOICE IN COOLANT IMPORTATION THROUGH TOTAL COSTS MINIMIZATION: A CASE STUDY

    Directory of Open Access Journals (Sweden)

    Marcela de Souza Leite

    2016-07-01

    Full Text Available Transportation plays a very significant role when it comes to the costs of a company representing on average 60% of logistics costs, so its management is very important for any company. The transportation modal choice is one of the most important transportation decisions. The purpose of this article is to select the transportation mode which is able to minimize total costs, and consistent with the objectives of customer service on the coolant import, which is used in plasma cutting machines. With the installation of a distribution center in Brazil and the professionalization of the logistics department of the company, it was decided to re-evaluate the transportation mode previously chosen to import some items. To determine the best mode of transportation was used basic compensation costs, in other words the cost compensation of using the shuttle service to the indirect cost of inventory related to the modal performance. Through the study, it was possible to observe it may be possible to save up to 73% on the coolant international transportation by changing the transportation mode used by the company.

  4. Effect of internal coolant crossflow orientation on the discharge coefficient of shaped film-cooling holes

    Energy Technology Data Exchange (ETDEWEB)

    Gritsch, M.; Saumweber, C.; Schulz, A.; Wittig, S.; Sharp, E.

    2000-01-01

    Discharge coefficients of three film-cooling hole geometries are presented over a wide range of engine like conditions. The hole geometries comprise a cylindrical hole and two holes with a diffuser-shaped exit portion (a fanshaped and a laidback fanshaped hole). For all three hole geometries the hole axis was inclined 30 deg with respect to the direction of the external (hot gas) flow. The flow conditions considered were the hot gas crossflow Mach number (up to 0.6), the coolant crossflow Mach number (up to 0.6) and the pressure ratio across the hole (up to 2). The effect of internal crossflow approach direction, perpendicular or parallel to the main flow direction, is particularly addressed in the present study. Comparison is made of the results for a parallel and perpendicular orientation, showing that the coolant crossflow orientation has a strong impact on the discharge behavior of the different hole geometries. The discharge coefficients were found to strongly depend on both hole geometry and crossflow conditions. Furthermore, the effects of internal and external crossflow on the discharge coefficients were described by means of correlations used to derive a predicting scheme for discharge coefficients. A comparison between predictions and measurements reveals the capability of the method proposed.

  5. Performance of water and diluted ethylene glycol as coolants for electronic cooling

    Directory of Open Access Journals (Sweden)

    M. Gayatri,

    2015-05-01

    Full Text Available As the number of transistors increases with new generation of microprocessor chips, the power draw and heat load to dissipate during operation increases. As a result of increasing the heat loads and heat fluxes the Conventional cooling technologies such as fan, heat sinks are unable to absorb and heat transfer excess heat dissipated by these new microprocessor. So, new technologies are needed to improve the heat removal capacity. In the present work single phase liquid cooling system with mini channel is analyzed and experimentally investigated. Mini channels are chosen as to provide higher heat transfer co-efficient than conventional channel. Copper pipes of 0.36 mm diameter are taken to fabricate heat sink and heat exchanger. A pump is used to circulate the fluid through heat sink and heat exchanger. A solid heated aluminium block to simulate heat generated electronic component is used and electrical input is supplied to the heated aluminium block and cooling system is placed over the heated block. The performance of the cooling system is analyzed from the experimental data obtained. It is experimentally observed that the mini channel liquid cooling system with water as a coolant has better performance than diluted ethylene glycol as coolant at different flow rates. The surface temperature of the heated aluminium block with convective heat transfer co-efficient is observed

  6. In-vessel ITER tubing failure rates for selected materials and coolants

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Cadwallader, L.C. [EG& G Idaho Inc., Idaho Falls, ID (United States)

    1994-03-01

    Several materials have been suggested for fabrication of ITER in-vessel coolant tubing: beryllium, copper, Inconel, niobium, stainless steel, titanium, and vanadium. This report generates failure rates for the materials to identify the best performer from an operational safety and availability perspective. Coolant types considered in this report are helium gas, liquid lithium, liquid sodium, and water. Failure rates for the materials are generated by including the influence of ITER`s operating environment and anticipated tubing failure mechanisms with industrial operating experience failure rates. The analyses define tubing failure mechanisms for ITER as: intergranular attack, flow erosion, helium induced swelling, hydrogen damage, neutron irradiation embrittlement, cyclic fatigue, and thermal cycling. K-factors, multipliers, are developed to model each failure mechanism and are applied to industrial operating experience failure rates to generate tubing failure rates for ITER. The generated failure rates identify the best performer by its expected reliability. With an average leakage failure rate of 3.1e-10(m-hr){sup {minus}1}and an average rupture failure rate of 3.1e-11(m-hr){sup {minus}1}, titanium proved to be the best performer of the tubing materials. The failure rates generated in this report are intended to serve as comparison references for design safety and optimization studies. Actual material testing and analyses are required to validate the failure rates.

  7. Design of coolant distribution system (CDS) for ITER PF AC/DC converter

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Bin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Zhiquan, E-mail: zhquansong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fu, Peng; Xu, Xuesong; Li, Chuan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Min; Dong, Lin [China International Nuclear Fusion Energy Program Execution Center, Beijing 100862 (China)

    2016-10-15

    Highlights: • System process and arrangement has been proposed to meet the multiple requirements from the converter system. • Thermal hydraulic analysis model has been developed to size and predict the system operation behavior. • Prototype test has been performed to validate the proposed design methodology. - Abstract: The Poloidal Field (PF) converter unit, playing an essential role in the plasma shape and position control in vertical and horizontal direction, which is an important part of ITER power supply system. As an important subsystem of the converter unit, the coolant distribution system has the function to distribute the cooling water from ITER component cooling water system (CCWS) to its main components at the required flow rate, pressure and temperature. This paper presents the thermal hydraulic design of coolant distribution system for the ITER PF converter unit. Different operational requirements of the PF converter unit regarding flow rate, temperature and pressure have been analyzed to design the system process and arrangement. A thermal-hydraulic analysis model has been built to size the system and predict the flow rate and temperature distribution of the system under the normal operation. Based on the system thermal-hydraulic analysis results, the system pressure profile has been plotted to evaluate the pressure behavior along each client flow path. A CDS prototype for the ITER PF converter has been constructed and some experiments have been performed on it. A good agreement of the flow distribution and temperature behavior between the simulated and test results validate the proposed design methodology.

  8. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  9. Determination of the {sup 129}I in primary coolant of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ke Chon; Park, Yong Joon; Song, Kyu Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-02-15

    Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as {sup 129}I is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for {sup 129}I as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding {sup 125}I as a radio-isotopic tracer (t1/2 = 60.14 d) to the simulation sample, in order to measure the activity concentration of {sup 129}I in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, Clform) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of {sup 129}I was examined, as was the effect of {sup 3}H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous {sup 3}H presence was found with activity concentrations of {sup 3}H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 {mu}g/mL.

  10. Numerical Investigation of Urea Freezing and Melting Characteristics Using Coolant Heater

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Yeop; Kim, Nam Il; Kim, Man Young [Chounbuk Nat' l Univ., Jeonju (Korea, Republic of); Park, Yun Beom [Jeju College of Technology, Jeju (Korea, Republic of)

    2013-08-15

    UREA-SCR technology is known as one of the powerful NOx reduction systems for vehicles as well as stationary applications. For its consistent and reliable operation in vehicle applications, however, the freezing and melting of the urea solution in cold environments have to be resolved. In this study, therefore, a numerical study of three-dimensional unsteady problems was analyzed to understand the urea freezing and heating phenomena and heat transfer characteristics in terms of urea liquid volume fraction, temperature profiles, and phase change behavior in urea solutions with time by using the commercial software Fluent 6.3. As a result, it was found that the freezing phenomenon proceeds with a phase change from the tank wall to the center, whereas the melting phenomenon occurs faster in the upper part of the storage tank by natural convection and in the adjacent part of the coolant pipe than in other parts. Furthermore, approximately 190s were required to obtain 1a of urea solution using a 4-coiled coolant heater under conditions of 70 .deg. C and 200 L/h.

  11. System Assessment of Carbon Dioxide Used as Gas Oxidant and Coolant in Vanadium-Extraction Converter

    Science.gov (United States)

    Du, Wei Tong; Wang, Yu; Liang, Xiao Ping

    2017-07-01

    With the aim of reducing carbon dioxide (CO2) emissions and of using waste resources in steel plants, the use of CO2 as a gas oxidant and coolant in the converter to increase productivity and energy efficiency was investigated in this study. Experiments were performed in combination with thermodynamic theory on vanadium-extraction with CO2 and oxygen (O2) mixed injections. The results indicate that the temperature of the hot metal bath decreased as the amount of CO2 introduced into O2 increased. At an injection of 85 vol.% O2 and 15 vol.% CO2, approximately 12% of additional carbon was retained in the hot metal. Moreover, the content of vanadium trioxide in the slag was higher. In addition, the O2 consumption per ton of hot metal was reduced by 8.5% and additional chemical energy was recovered by the controlled injection of CO2 into the converter. Therefore, using CO2 as a gas coolant was conducive to vanadium extraction, and O2 consumption was reduced.

  12. System Assessment of Carbon Dioxide Used as Gas Oxidant and Coolant in Vanadium-Extraction Converter

    Science.gov (United States)

    Du, Wei Tong; Wang, Yu; Liang, Xiao Ping

    2017-10-01

    With the aim of reducing carbon dioxide (CO2) emissions and of using waste resources in steel plants, the use of CO2 as a gas oxidant and coolant in the converter to increase productivity and energy efficiency was investigated in this study. Experiments were performed in combination with thermodynamic theory on vanadium-extraction with CO2 and oxygen (O2) mixed injections. The results indicate that the temperature of the hot metal bath decreased as the amount of CO2 introduced into O2 increased. At an injection of 85 vol.% O2 and 15 vol.% CO2, approximately 12% of additional carbon was retained in the hot metal. Moreover, the content of vanadium trioxide in the slag was higher. In addition, the O2 consumption per ton of hot metal was reduced by 8.5% and additional chemical energy was recovered by the controlled injection of CO2 into the converter. Therefore, using CO2 as a gas coolant was conducive to vanadium extraction, and O2 consumption was reduced.

  13. Experimental and analytical studies of melt jet-coolant interactions: a synthesis

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Bui, V.A.; Nourgaliev, R.R.; Green, J.A.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-01-01

    Instability and fragmentation of a core melt jet in water have been actively studied during the past ten years. Several models, and a few computer codes, have been developed. However, there are, still, large uncertainties, both, in interpreting experimental results and in predicting reactor-scale processes. Steam explosion and debris coolability, as reactor safety issues, are related to the jet fragmentation process. A better understanding of the physics of jet instability and fragmentation is crucial for assessments of fuel-coolant interactions (FCIs). This paper presents research, conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning molten jet-coolant interactions, as a precursor for premixing. First, observations were obtained from scoping experiments with simulant fluids. Second, the linear perturbation method was extended and applied to analyze the interfacial-instability characteristics. Third, two innovative approachs to CFD modeling of jet fragmentation were developed and employed for analysis. The focus of the studies was placed on (a) identifying potential factors, which may affect the jet instability, (b) determining the scaling laws, and (c) predicting the jet behavior for severe accidents conditions. In particular, the effects of melt physical properties, and the thermal hydraulics of the mixing zone, on jet fragmentation were investigated. Finally, with the insights gained from a synthesis of the experimental results and analysis results, a new phenomenological concept, named `macrointeractions concept of jet fragmentation` is proposed. (author)

  14. Heat Exchanger Can Assembly for Provision of Helium Coolant Streams for Cryomodule Testing below 2K

    Science.gov (United States)

    Smith, E. N.; Eichhorn, R.; Quigley, P.; Sabol, D.; Shore, C.; Widger, D.

    2017-02-01

    A series of heat exchanger can (HXC) assemblies have been designed, constructed and built to utilize existing 4.2 K liquefaction and compressor capabilities to provide helium gas coolant streams of 80 K, 4.5 K, and liquid from 1.6 to 2.0 K for operating cryomodules containing from one to six superconducting RF cavities built for an energy recovery linear accelerator. Designs for the largest assemblies required up to 100 W of cooling at 1.8 K with precise temperature control, especially during cool-down, and up to 2000 W at 80 K (with a 40 K temperature rise). A novel feature of these assemblies was the use of relatively inexpensive brazed stainless steel plate heat exchangers intended for room-temperature operation with water or oil, but which in practice worked well at cryogenic temperatures. The choice of operating temperatures/pressures were to provide single-phase helium flow for better control of coolant distribution in the 80 K and 4.5 K streams, to take advantage of locally elevated heat capacity near the critical point for the 4.5 K stream, and in the region below 2 K to get the best possible Q from the niobium cavities under test.

  15. High temperature storage loop :

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.

    2013-07-01

    A three year plan for thermal energy storage (TES) research was created at Sandia National Laboratories in the spring of 2012. This plan included a strategic goal of providing test capability for Sandia and for the nation in which to evaluate high temperature storage (>650ÀC) technology. The plan was to scope, design, and build a flow loop that would be compatible with a multitude of high temperature heat transfer/storage fluids. The High Temperature Storage Loop (HTSL) would be reconfigurable so that it was useful for not only storage testing, but also for high temperature receiver testing and high efficiency power cycle testing as well. In that way, HTSL was part of a much larger strategy for Sandia to provide a research and testing platform that would be integral for the evaluation of individual technologies funded under the SunShot program. DOEs SunShot program seeks to reduce the price of solar technologies to 6/kWhr to be cost competitive with carbon-based fuels. The HTSL project sought to provide evaluation capability for these SunShot supported technologies. This report includes the scoping, design, and budgetary costing aspects of this effort

  16. High temperature storage loop :

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.

    2013-07-01

    A three year plan for thermal energy storage (TES) research was created at Sandia National Laboratories in the spring of 2012. This plan included a strategic goal of providing test capability for Sandia and for the nation in which to evaluate high temperature storage (>650ÀC) technology. The plan was to scope, design, and build a flow loop that would be compatible with a multitude of high temperature heat transfer/storage fluids. The High Temperature Storage Loop (HTSL) would be reconfigurable so that it was useful for not only storage testing, but also for high temperature receiver testing and high efficiency power cycle testing as well. In that way, HTSL was part of a much larger strategy for Sandia to provide a research and testing platform that would be integral for the evaluation of individual technologies funded under the SunShot program. DOEs SunShot program seeks to reduce the price of solar technologies to 6/kWhr to be cost competitive with carbon-based fuels. The HTSL project sought to provide evaluation capability for these SunShot supported technologies. This report includes the scoping, design, and budgetary costing aspects of this effort

  17. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to

  18. Methodology for determining of the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V. [JSC ' Atomtechenergo' , Filial ' Novovoronezhatomtechenergo' , Novovorenezh (Russian Federation); Ryasny, Sergei I. [JSC ' Atomtechenergo' , Mytishi (Russian Federation)

    2016-09-15

    At WWER-1000 NPPs, as well as at PWR NPPs, there is a problem of determining the correct weighted mean coolant temperature in the primary circuit hot legs based on the measuring channels information. The problem is caused by the coolant temperature stratification. The technical documentation for engineering support and maintenance of I and C systems does not provide any regulatory guidelines to consider this effect. Therefore, it is very important to represent a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of the WWER-1000 reactor plants. The given paper presents the basic preconditions and approaches applied during the methodology development. They were worked out on the basis of the executed numerical and experimental research taking into account the analysis of the extensive material obtained by the authors from full-scale tests during the commissioning of WWER-1000 power units, as well as operational data obtained from several power units with different fuel loadings.

  19. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  20. Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes. 1. Investigation on velocity ratio between the coolant pipes

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-02-01

    This report presents numerical results on thermal striping characteristics at a tee junction of LMFBR coolant pipe, carried out using a direct numerical simulation code DINUS-3. In the numerical investigations, it was considered a tee junction system consisted of a main pipe (1.33 cm{sup I.D.}) with a 90deg elbow and a branch pipe having same inner diameter to the main pipe, and five velocity ratio conditions between both the pipes, i.e., (V{sub main}/V{sub branch}) = 0.25; 0.5; 1.0; 2.0 and 4.0. From the numerical investigations the following characteristics were obtained: (1) Temperature fluctuations in the downstream region of the tee junction were formulated by lower frequency components (<7.0 Hz) due to the interactions between main pipe flows and jet flows from the branch pipe, and higher frequency components (>10.0 Hz) generated by the vortex released frequency from the outer edge of the branch pipe jet flows. (2) On the top plane of the main pipe, peak values of the temperature fluctuation amplitude was decreased with increasing flow velocity in the main pipe, and its position was shifted to downstream direction of the main pipe by the increase of the main pipe flow velocity. (3) On the bottom plane of the main pipe, contrary to (2), peak values of the temperature fluctuation amplitude was increased with increasing flow velocity in the main pipe. (author)

  1. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde; Simulacion de escenarios de accidente severo tipo perdida de refrigerante (Loca) pequeno, con el codigo MELCOR version 2.1 para la central nucleo-electrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft{sup 2} (4.6 cm{sup 2}), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  2. Condition Monitoring of Control Loops

    OpenAIRE

    Horch, Alexander

    2000-01-01

    The main concern of this work is the development of methodsfor automatic condition monitoring of control loops withapplication to the process industry. By condition monitoringboth detection and diagnosis of malfunctioning control loops isunderstood, using normal operating data and a minimum amount ofprocess knowledge. The use of indices for quantifying loop performance is dealtwith in the first part of the thesis. The starting point is anindex proposed by Harris (1989). This index has been mo...

  3. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  4. High Temperature Fluoride Salt Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.

  5. Loop Heat Pipe Startup Behaviors

    Science.gov (United States)

    Ku, Jentung

    2016-01-01

    A loop heat pipe must start successfully before it can commence its service. The startup transient represents one of the most complex phenomena in the loop heat pipe operation. This paper discusses various aspects of loop heat pipe startup behaviors. Topics include the four startup scenarios, the initial fluid distribution between the evaporator and reservoir that determines the startup scenario, factors that affect the fluid distribution between the evaporator and reservoir, difficulties encountered during the low power startup, and methods to enhance the startup success. Also addressed are the pressure spike and pressure surge during the startup transient, and repeated cycles of loop startup and shutdown under certain conditions.

  6. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    Science.gov (United States)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  7. Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant

    Science.gov (United States)

    Belyaev, I. A.; Genin, L. G.; Krylov, S. G.; Novikov, A. O.; Razuvanov, N. G.; Sviridov, V. G.

    2015-09-01

    The aim of this experimental investigation is to obtain information on the temperature fields and heat transfer coefficients during flow of liquid-metal coolant in models simulating an elementary cell in the core of a liquid heavy metal cooled fast-neutron reactor. Two design versions for spacing fuel rods in the reactor core were considered. In the first version, the fuel rods were spaced apart from one another using helical wire wound on the fuel rod external surface, and in the second version spacer grids were used for the same purpose. The experiments were carried out on the mercury loop available at the Moscow Power Engineering Institute National Research University's Chair of Engineering Thermal Physics. Two experimental sections simulating an elementary cell for each of the fuel rod spacing versions were fabricated. The temperature fields were investigated using a dedicated hinged probe that allows temperature to be measured at any point of the studied channel cross section. The heat-transfer coefficients were determined using the wall temperature values obtained at the moment when the probe thermocouple tail end touched the channel wall. Such method of determining the wall temperature makes it possible to alleviate errors that are unavoidable in case of measuring the wall temperature using thermocouples placed in slots milled in the wall. In carrying out the experiments, an automated system of scientific research was applied, which allows a large body of data to be obtained within a short period of time. The experimental investigations in the first test section were carried out at Re = 8700, and in the second one, at five values of Reynolds number. Information about temperature fields was obtained by statistically processing the array of sampled probe thermocouple indications at 300 points in the experimental channel cross section. Reach material has been obtained for verifying the codes used for calculating velocity and temperature fields in channels with

  8. Dynamic PID loop control

    Energy Technology Data Exchange (ETDEWEB)

    Pei, L.; Klebaner, A.; Theilacker, J.; Soyars, W.; Martinez, A.; Bossert, R.; DeGraff, B.; Darve, C.; /Fermilab

    2011-06-01

    The Horizontal Test Stand (HTS) SRF Cavity and Cryomodule 1 (CM1) of eight 9-cell, 1.3GHz SRF cavities are operating at Fermilab. For the cryogenic control system, how to hold liquid level constant in the cryostat by regulation of its Joule-Thompson JT-valve is very important after cryostat cool down to 2.0 K. The 72-cell cryostat liquid level response generally takes a long time delay after regulating its JT-valve; therefore, typical PID control loop should result in some cryostat parameter oscillations. This paper presents a type of PID parameter self-optimal and Time-Delay control method used to reduce cryogenic system parameters oscillation.

  9. Inductance loop and partial

    CERN Document Server

    Paul, Clayton R

    2010-01-01

    "Inductance is an unprecedented text, thoroughly discussing "loop" inductance as well as the increasingly important "partial" inductance. These concepts and their proper calculation are crucial in designing modern high-speed digital systems. World-renowned leader in electromagnetics Clayton Paul provides the knowledge and tools necessary to understand and calculate inductance." "With the present and increasing emphasis on high-speed digital systems and high-frequency analog systems, it is imperative that system designers develop an intimate understanding of the concepts and methods in this book. Inductance is a much-needed textbook designed for senior and graduate-level engineering students, as well as a hands-on guide for working engineers and professionals engaged in the design of high-speed digital and high-frequency analog systems."--Jacket.

  10. Dynamic PID loop control

    CERN Document Server

    Pei, L; Theilacker, J; Soyars, W; Martinez, A; Bossert, R; DeGraff, B; Darve, C

    2012-01-01

    The Horizontal Test Stand (HTS) SRF Cavity and Cryomodule 1 (CM1) of eight 9-cell, 1.3GHz SRF cavities are operating at Fermilab. For the cryogenic control system, how to hold liquid level constant in the cryostat by regulation of its Joule-Thompson JT-valve is very important after cryostat cool down to 2.0 K. The 72-cell cryostat liquid level response generally takes a long time delay after regulating its JT-valve; therefore, typical PID control loop should result in some cryostat parameter oscillations. This paper presents a type of PID parameter self-optimal and Time-Delay control method used to reduce cryogenic system parameters' oscillation.

  11. Vortex loops and Majoranas

    Energy Technology Data Exchange (ETDEWEB)

    Chesi, Stefano [Department of Physics, McGill University, Montreal, Quebec H3A 2T8 (Canada); CEMS, RIKEN, Wako, Saitama 351-0198 (Japan); Jaffe, Arthur [Harvard University, Cambridge, Massachusetts 02138 (United States); Department of Physics, University of Basel, Basel (Switzerland); Institute for Theoretical Physics, ETH Zürich, Zürich (Switzerland); Loss, Daniel [CEMS, RIKEN, Wako, Saitama 351-0198 (Japan); Department of Physics, University of Basel, Basel (Switzerland); Pedrocchi, Fabio L. [Department of Physics, University of Basel, Basel (Switzerland)

    2013-11-15

    We investigate the role that vortex loops play in characterizing eigenstates of interacting Majoranas. We give some general results and then focus on ladder Hamiltonian examples as a test of further ideas. Two methods yield exact results: (i) A mapping of certain spin Hamiltonians to quartic interactions of Majoranas shows that the spectra of these two examples coincide. (ii) In cases with reflection-symmetric Hamiltonians, we use reflection positivity for Majoranas to characterize vortices in the ground states. Two additional methods suggest wider applicability of these results: (iii) Numerical evidence suggests similar behavior for certain systems without reflection symmetry. (iv) A perturbative analysis also suggests similar behavior without the assumption of reflection symmetry.

  12. Loop Quantum Gravity

    Directory of Open Access Journals (Sweden)

    Rovelli Carlo

    2008-07-01

    Full Text Available The problem of describing the quantum behavior of gravity, and thus understanding quantum spacetime, is still open. Loop quantum gravity is a well-developed approach to this problem. It is a mathematically well-defined background-independent quantization of general relativity, with its conventional matter couplings. Today research in loop quantum gravity forms a vast area, ranging from mathematical foundations to physical applications. Among the most significant results obtained so far are: (i The computation of the spectra of geometrical quantities such as area and volume, which yield tentative quantitative predictions for Planck-scale physics. (ii A physical picture of the microstructure of quantum spacetime, characterized by Planck-scale discreteness. Discreteness emerges as a standard quantum effect from the discrete spectra, and provides a mathematical realization of Wheeler’s “spacetime foam” intuition. (iii Control of spacetime singularities, such as those in the interior of black holes and the cosmological one. This, in particular, has opened up the possibility of a theoretical investigation into the very early universe and the spacetime regions beyond the Big Bang. (iv A derivation of the Bekenstein–Hawking black-hole entropy. (v Low-energy calculations, yielding n-point functions well defined in a background-independent context. The theory is at the roots of, or strictly related to, a number of formalisms that have been developed for describing background-independent quantum field theory, such as spin foams, group field theory, causal spin networks, and others. I give here a general overview of ideas, techniques, results and open problems of this candidate theory of quantum gravity, and a guide to the relevant literature.

  13. Study on the effect of the impeller and diffuser blade number on reactor coolant pump performances

    Science.gov (United States)

    Long, Y.; Yin, J. L.; Wang, D. Z.; Li, T. B.

    2016-05-01

    In this paper, CFD approach was employed to study how the blade number of impeller and diffuser influences reactor coolant pump performances. The three-dimensional pump internal flow channel was modelled by pro/E software, Reynolds-averaged Naiver-Stokes equations with the k-ε turbulence model were solved by the computational fluid dynamics software CFX. By post-processing on the numerical results, the performance curves of reactor coolant pump were obtained. The results are as follows, with the blade number of the impeller increasing, the head of the pump with different diffuser universally increases in the 8Q n∼1.2Q n conditions, and at different blade number of the diffuser, the head increases with the blade number of the impeller increasing. In 1.0Q n condition, when the blades number combination of impeller and diffuser chooses 4+16, 7+14 and 6+18, the head curves exist singular points. In 1.2Q n condition, the head curve still exists singular point in 6+18. With the blade number of the impeller increasing, the efficiency of the pump with different diffuser universally decreases in the 0.8Q n and 1.0Q n conditions, but in 1.2Q n condition, the efficiency of the pump with different diffuser universally increases. In 1.0Q n condition, the impellers of 4 and 5 blades are better. When the blade number combination of impeller and diffuser choose 4+11, 4+17, 4+18, 5+12, 5+17 and 5+18, the efficiencies relatively have higher values. With the blade number of the impeller increasing, the hydraulic shaft power of the pump with different diffuser universally increases in the 0.8Q n∼1.2Q n conditions, and with the blade number of the diffuser increasing, the power of different impeller overall has small fluctuation, but tends to be uniform. This means the increase of the diffuser blade number has less influence on shaft power.The influence on the head and flow by the matching relationship of the blades number between impeller and diffuser is very complicated, which

  14. Simulation of Heat Transfer to the Gas Coolant with Low Prandtl Number Value

    Directory of Open Access Journals (Sweden)

    T. N. Kulikova

    2015-01-01

    Full Text Available The work concerns the simulating peculiarities of heat transfer to the gas coolants with low values of the Prandtl number, in particular, to the binary mixtures of inert gases.The paper presents simulation results of heat transfer to the fully established flow of a helium-xenon mixture in the round tube of 6 mm in diameter with the boundary condition of the second kind. It considers a flow of three helium-xenon mixtures with different helium content and molecular Prandtl numbers within the range 0.239–0.322 and with Reynolds numbers ranged from 10000 to 50000. During numerical simulation a temperature factor changed from 1.034 to 1.061. CFD-code STAR-CCM+ that is designed for solving a wide range of problems of hydrodynamics, heat transfer and stress was used as the primary software.The applicability of the five models for the turbulent Prandtl number is examined. It is shown that the choice of the model has a significant influence on the heat transfer coefficient. The paper presents structural characteristics of the flow in the wall region. It estimates a thermal stabilization section to be approximately as long as 30 diameters of tube.Simulation results are compared with the known data on heat transfer to gas coolants with low values of the Prandtl number. It is shown that V2F low-Reynolds number -ε turbulence model with an approximation for the turbulent Prandtl number used according Kays-CrawfordWeigand gives the best compliance with the results predicted by relationships of Kays W.M. and Petukhov B.S. The approximating correlation summarizes a set of simulation results.Application of the work results is reasonable when conducting the numerical simulation of heat transfer to binary gas mixtures in channels of different forms. The presented approximating correlation allows rapid estimate of heat transfer coefficients to the gas coolants with a low value of the molecular Prandl number within the investigated range with a flow through the

  15. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  16. Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

    Energy Technology Data Exchange (ETDEWEB)

    Gerard, R.; Malekian, C.; Meessen, O. [Tractebel Energy Engineering, Brussels (Belgium)

    1997-04-01

    The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

  17. Phenomenology of loop quantum cosmology

    CERN Document Server

    Sakellariadou, Mairi

    2010-01-01

    After introducing the basic ingredients of Loop Quantum Cosmology, I will briefly discuss some of its phenomenological aspects. Those can give some useful insight about the full Loop Quantum Gravity theory and provide an answer to some long-standing questions in early universe cosmology.

  18. RCD+: Fast loop modeling server.

    Science.gov (United States)

    López-Blanco, José Ramón; Canosa-Valls, Alejandro Jesús; Li, Yaohang; Chacón, Pablo

    2016-07-08

    Modeling loops is a critical and challenging step in protein modeling and prediction. We have developed a quick online service (http://rcd.chaconlab.org) for ab initio loop modeling combining a coarse-grained conformational search with a full-atom refinement. Our original Random Coordinate Descent (RCD) loop closure algorithm has been greatly improved to enrich the sampling distribution towards near-native conformations. These improvements include a new workflow optimization, MPI-parallelization and fast backbone angle sampling based on neighbor-dependent Ramachandran probability distributions. The server starts by efficiently searching the vast conformational space from only the loop sequence information and the environment atomic coordinates. The generated closed loop models are subsequently ranked using a fast distance-orientation dependent energy filter. Top ranked loops are refined with the Rosetta energy function to obtain accurate all-atom predictions that can be interactively inspected in an user-friendly web interface. Using standard benchmarks, the average root mean squared deviation (RMSD) is 0.8 and 1.4 Å for 8 and 12 residues loops, respectively, in the challenging modeling scenario in where the side chains of the loop environment are fully remodeled. These results are not only very competitive compared to those obtained with public state of the art methods, but also they are obtained ∼10-fold faster. © The Author(s) 2016. Published by Oxford University Press on behalf of Nucleic Acids Research.

  19. Improved code-tracking loop

    Science.gov (United States)

    Laflame, D. T.

    1980-01-01

    Delay-locked loop tracks pseudonoise codes without introducing dc timing errors, because it is not sensitive to gain imbalance between signal processing arms. "Early" and "late" reference codes pass in combined form through both arms, and each arm acts on both codes. Circuit accomodates 1 dB weaker input signals with tracking ability equal to that of tau-dither loops.

  20. Loop groups and noncommutative geometry

    CERN Document Server

    Carpi, Sebastiano

    2015-01-01

    We describe the representation theory of loop groups in terms of K-theory and noncommutative geometry. This is done by constructing suitable spectral triples associated with the level l projective unitary positive-energy representations of any given loop group LG. The construction is based on certain supersymmetric conformal field theory models associated with LG.

  1. Brane Couplings from Bulk Loops

    OpenAIRE

    Georgi, Howard; Grant, Aaron K.; Hailu, Girma

    2000-01-01

    We compute loop corrections to the effective action of a field theory on a five-dimensional $S_1/Z_2$ orbifold. We find that the quantum loop effects of interactions in the bulk produce infinite contributions that require renormalization by four-dimensional couplings on the orbifold fixed planes. Thus bulk couplings give rise to renormalization group running of brane couplings.

  2. Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes. 4. Investigation on second-order moments in coolant mixing region

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2000-02-01

    This report presents numerical results on thermal striping characteristics at a tee junction of LMFBR coolant pipe, carried out using a direct numerical simulation code DINUS-3. In the numerical investigations, it was considered a tee junction system consisted of a main pipe (1.33 cm{sup I.D.}) with a 90deg elbow and a brunch pipe, and four parameters, i.e., (1) diameter ratio {alpha} between both the pipes, (2) flow velocity ratio {beta} between both the pipes, (3) angle {gamma} between both the pipes, and (4) Reynolds number Re. From the numerical investigations, the following characteristics were obtained: (1) According to the decreasing of the diameter ratio, significant area of second-order moments was expanded in the fixed condition of {beta}=1.0. (2) Significant second-order moments area was expanded for the increasing of the flow velocity ratio {beta} specified by varying of the main pipe velocity in the case of a {alpha}=1.0 constant condition. On the other hand, the area was expanded for the decreasing of the velocity ratio {beta} defined by varying of the branch pipe velocity in the case of a {alpha}=1.0 constant condition. (3) Maximum second-order moments values were generated in the case of {gamma}=180deg due to the influence of interactions between main pipe flows and jet flows from the branch pipe. (4) According to the increase of Reynolds number, significant area of second-order moments was expanded due to the activation of turbulence mixing in the main pipe. (author)

  3. Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, Ashwini K., E-mail: ashwinikumaryadav@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ravi, E-mail: ravikfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Chatterjee, B., E-mail: barun@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, Akhilesh, E-mail: akhilfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Mukhopadhyay, D., E-mail: dmukho@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2013-02-15

    Highlights: ► Circumferential temperature gradient of PT for asymmetric heat-up was 440 °C. ► At 2 MPa ballooning initiated at 450 °C and with strain rate of 0.0277%/s. ► At 4 MPa ballooning initiated at 390 °C and with strain rate of 0.0305%/s. ► At 4 MPa, PT ruptured under uneven strain and steep temperature gradient. ► Integrity of PT depends on internal pressure and magnitude of decay power. -- Abstract: During postulated small break loss of coolant accident (SBLOCA) for Pressurised Heavy Water Reactors (PHWRs) as well as for postulated SBLOCA coincident with loss of ECCS, a stratified flow condition can arise in the coolant channels as the gravitational force dominates over the low inertial flow arising from small break flow. A Station Blackout condition without operator intervention can also lead to stratified flow condition during a slow channel boil-off condition. For all these conditions the pressure remains high and under stratified flow condition, the horizontal fuel bundles experience different heat transfer environments with respect to the stratified flow level. This causes the bundle upper portion to get heated up higher as compared to the submerged portion. This kind of asymmetrical heating of the bundle is having a direct bearing on the circumferential temperature gradient of pressure tube (PT) component of the coolant channel. The integrity of the PT is important under normal conditions as well as at different accident loading conditions as this component houses the fuel bundles and serves as a coolant pressure boundary of the reactors. An assessment of PT is required with respect to different accident loading conditions. The present investigation aims to study thermo-mechanical behaviour of PT (Zr, 2.5 wt% Nb) under a stratified flow condition under different internal pressures. The component is subjected to an asymmetrical heat-up conditions as expected during the said situation under different pressure conditions which varies from 2

  4. Effectiveness of non-volatile falling film absorbers with solution and coolant in counter-flow

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D.S. [Austrian Institute of Technology, Dept. Energy, Giefinggasse 2, 1210 Vienna (Austria); Infante Ferreira, C.A. [Delft University of Technology, Engineering Thermodynamics, Leeghwaterstraat 44, 2628 CA Delft (Netherlands)

    2010-01-15

    Effectiveness equations are developed for non-volatile falling film absorbers with solution and coolant in counter-flow. It is shown how mixture thermodynamics and film theory can be used to simplify the problem to give eigenvalue solutions for temperature and concentration profiles and how heat and mass transfer effectiveness equations can be derived from such solutions. The results indicate that the transfer process in an absorber is driven by two driving forces, i.e. the difference between bulk solution and cooling water temperatures and the initial deviation of bulk solution from its equilibrium state. Asymptotic effectiveness equations are derived for a few limiting cases to show that they approach their counterparts in single-phase heat transfer and isothermal absorption processes. (author)

  5. Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

    Science.gov (United States)

    Leenaers, A.; Joppen, F.; Van den Berghe, S.

    2009-10-01

    In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% 235U enriched UAl x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% 235U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE. Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900-950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.

  6. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kao, S.P.; Chang, S.K.; Huang, H.C. [Nuclear Training Branch, Northeast Utilities, Waterford, CT (United States)

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

  7. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  8. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    Energy Technology Data Exchange (ETDEWEB)

    Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  9. Safety analysis of the US dual coolant liquid lead lithium ITER test blanket module

    Science.gov (United States)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2007-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  10. Chemical and radiolytical characterization of some perfluorocarbon fluids used as coolants for LHC experiments

    CERN Document Server

    Battistin, M; Setnescu, R; Teissandier, B; CERN. Geneva. TS Department

    2006-01-01

    Perfluorocarbon fluids, - mainly C6F14 - used as coolants within High Energy Physics Detectors in the Large Hadrons Collider (LHC) at CERN, were characterized by applying mainly the following methods: GC, FT-IR and UV-Vis. The aim of this work was the quality control, the identification and the quantification of different impurities which could increase the radiation sensitivity of these fluids. Thus, the presence of H containing molecules within perfluorocarbons strongly influences the appearance of hydrofluoric acid during their irradiation. The procedures settled-up in this work are sensitive to the presence of such impurities and would be used for the analyses of the received perfluorocarbon fluids as well as to assess the radiation induced modifications and the efficiency of their purification treatments.

  11. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  12. Effect of the Shrink Fit and Mechanical Tolerance on Reactor Coolant Pump Flywheel Integrity Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Donghak [Korea KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Reactor coolant pump (RCP) flywheel should satisfy the RCP flywheel integrity criteria of the US NRC standard review plan (SRP) 5.4.1.1 and regulatory guide (RG) 1.14. Shrink-fit and rotational stresses should be calculated to evaluate the integrity. In this paper the effects of the shrink fit and mechanical tolerance on the RCP flywheel integrity evaluation are studied. The shrink fit should be determined by the joint release speed and the stresses in the flywheel will be increased by the shrink fit. The stress at the interface between the hub and the outer wheel shows the highest value. The effect of the mechanical tolerance should be considered for the stress evaluation. And the effect of the mechanical tolerance should be not considered to determine the joint release speed.

  13. Vibration signal analysis of main coolant pump flywheel based on Hilert-Huang transform

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Meiru; Xia, Hong; Sun, Lin; Li, Bin; Yang, Yang [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, College of Nuclear Science and Technology, Harbin Engineering University, Harbin (China)

    2015-03-15

    In this paper, a three-dimensional model for the dynamic analysis of a flywheel based on the finite element method is presented. The static structure analysis for the model provides stress and strain distribution cloud charts. The modal analysis provides the basis of dynamic analysis due to its ability to obtain the natural frequencies and the vibration made vectors of the first 10 orders. The results show the main faults are attrition and cracks, while also indicating the locations and patterns of faults. The harmonic response simulation was performed to gain the vibration response of the flywheel under operation. In this paper, we present a Hilberte-Huang transform (HHT) algorithm for flywheel vibration analysis. The simulation indicated that the proposed flywheel vibration signal analysis method performs well, which means that the method can lay the foundation for the detection and diagnosis in a reactor main coolant pump.

  14. System approach in the investigation of coolant parametrical oscillations in passive safety injection systems (PSIS)

    Energy Technology Data Exchange (ETDEWEB)

    Proskouriakov, K.N. [Moskovskij Ehnergeticheskij Inst., Moscow (Russian Federation)

    2001-07-01

    The use of thermal-hydraulic computer codes is an important part of the work programme for activities in the field of nuclear power plants (NPP) Safety Research as it will enable to define better the test configuration and parameter range extensions and to extrapolate the results of the small scale experiments towards full scale reactor applications. The CATHARE2, RELAP5, the WCOBRA/TRAC, and APROS codes are the estimate thermal hydraulic codes for the evaluation of large and small break loss of coolant accidents (LOCA). The relatively good agreement experimental data with the calculations have been presented. There was shown also some big mistakes in predicting distribution of flow when two phase are present. Model of parametrical oscillation (P.O.) worked out gives explanation for flow oscillations and indicates that the phenomenon of P.O. appears under certain combination of thermal-hydraulic parameters and structure of heat-removal system. (orig.)

  15. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  16. Liquid Cooling of Tractive Lithium Ion Batteries Pack with Nanofluids Coolant.

    Science.gov (United States)

    Li, Yang; Xie, Huaqing; Yu, Wei; Li, Jing

    2015-04-01

    The heat generated from tractive lithium ion batteries during discharge-charge process has great impacts on the performances of tractive lithium ion batteries pack. How to solve the thermal abuse in tractive lithium ion batteries pack becomes more and more urgent and important for future development of electrical vehicles. In this work, TiO2, ZnO and diamond nanofluids are prepared and utilized as coolants in indirect liquid cooling of tractive lithium ion batteries pack. The results show that nanofluids present superior cooling performance to that of pure fluids and the diamond nanofluid presents relatively excellent cooling abilities than that of TiO2 and ZnO nanofluids. During discharge process, the temperature distribution of batteries in batteries pack is uniform and stable, due to steady heat dissipation by indirect liquid cooling. It is expected that nanofluids could be considered as a potential alternative for indirect liquid cooling in electrical vehicles.

  17. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Heames, T.J. (Science Applications International Corp., Albuquerque, NM (USA)); Williams, D.A.; Johns, N.A.; Chown, N.M. (UKAEA Atomic Energy Establishment, Winfrith (UK)); Bixler, N.E.; Grimley, A.J. (Sandia National Labs., Albuquerque, NM (USA)); Wheatley, C.J. (UKAEA Safety and Reliability Directorate, Culcheth (UK))

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  18. CATHARE Multi-1D Modeling of Coolant Mixing in VVER-1000 for RIA Analysis

    Directory of Open Access Journals (Sweden)

    I. Spasov

    2010-01-01

    Full Text Available The paper presents validation results for multichannel vessel thermal-hydraulic models in CATHARE used in coupled 3D neutronic/thermal hydraulic calculations. The mixing is modeled with cross flows governed by local pressure drops. The test cases are from the OECD VVER-1000 coolant transient benchmark (V1000CT and include asymmetric vessel flow transients and main steam line break (MSLB transients. Plant data from flow mixing experiments are available for comparison. Sufficient mesh refinement with up to 24 sectors in the vessel is considered for acceptable resolution. The results demonstrate the applicability of such validated thermal-hydraulic models to MSLB scenarios involving thermal mixing, azimuthal flow rotation, and primary pump trip. An acceptable trade-off between accuracy and computational efficiency can be obtained.

  19. Development of manufacturing technology and fabrication of prototype for main coolant pump

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Koon Seok; Han, C.K.; Chei, J.M.; Chung, K.S.; Youn, M.H.; Shin, S.A.; Choi, D.J.; Kim, H.C. [HALLA Industrial Co., Ltd., Pusan (Korea)

    1999-03-01

    This study presents the development of the manufacturing technology for the Main Coolant Pump of the SMART. This report contains the followings; (1) Select axial type pump for the MCP (2) MCP is drived by squirrel-cage induction motor that consisted canned motor type. (3) MCP shaft has three horizontal and one vertical support bearings. (4) Design of several part of the MCP (5) Manufacturing of the performance test motor (6) Design and manufacturing of the speed sensor (7) Procedures for three-axial and five-axial M.C.T., Tig welding and Electron Beam Welding were developed. (8) Conceptional design of the MCP test facility for the performance test under operating conditions. (9) Results of standard weld test specimens according to the ASME section IX. (author). 21 refs., 35 figs., 10 tabs.

  20. Results and Observations of the Integral Loss-of-coolant Accident Test with Surface Modified Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun; Jung, Yang Il; Park, Jung Hwan; Kim, Hyun Gil; Yang, Jae Ho; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, integral loss-of-coolant accident (LOCA) test was carried for comprehensive understanding of phenomena such as ballooning, burst failures, and oxidation for the ATF cladding during a LOCA scenario. In this section some of the experimental procedure and technical details of apparatus are described. Highlight data obtained from simulated LOCA test is also presented. Cracks can be initiated at this brittle burst tip and will propagate rapidly though the ballooned region. Therefore, the flexural strength of the ruptured tubes mainly depends on the thickness of the load bearing Zr metal at the opposite side to the rupture opening. To improve the reliability and safety of existing Zr alloy fuel cladding under LWR accident conditions, a high temperature oxidation resistant layer was coated onto the surface of Zr alloy samples using various coating techniques. The rupture temperature of the coated tube was higher than that of the uncoated cladding.

  1. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  2. Coolant choice for the central beryllium pipe of the BESIII beam pipe

    Science.gov (United States)

    Zheng, Li-Fang; Wang, Li; Wu, Ping; Ji, Quan; Li, Xun-Feng; Liu, Jian-Ping

    2010-07-01

    In order to take away much more heat on the BESIII beam pipe to guarantee the normal particle detection, EDM-1 (oil No.1 for electric discharge machining), with good thermal and flow properties was selected as the candidate coolant for the central beryllium pipe of the BESIII beam pipe. Its cooling character was studied and dynamic corrosion experiment was undertaken to examine its corrosion on beryllium. The experiment results show that EDM-1 would corrode the beryllium 19.9 μm in the depth in 10 years, which is weak and can be neglected. Finite element simulation and experiment research were taken to check the cooling capacity of EDM-1. The results show that EDM-1 can meet the cooling requirement of the central beryllium pipe. Now EDM-1 is being used to cool the central beryllium pipe of the BESIII beam pipe.

  3. Automatic control of the lithium concentration of the reactor coolant in PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Long, A.; Bruere, X. [Framatome ANP, Paris (France); Cohen, J. [Electricite de France-DIS-CIPN, Marseille (France); Berger, M. [Electricite de France-DIS-SEPTEN, Villeurbanne (France)

    2002-07-01

    Given the specific operating mode of French units, observance of the lithium-boron diagram and consequently observance of reactor coolant pH is considered to be a priority relative to management of {sup 7}Li ({sup 7}Li recycling practices or prototypes). For this reason EDF and FRAMATOME-ANP have developed an automatic lithium hydroxide injection device, which serves to compensate in real time whenever the upper or lower limits of the lithium-boron diagram are exceeded and to prevent excursion at low pH. A prototype of this device is installed on unit N 2 of Tricastin NPP. The purpose of this document is to describe its principles and the main characteristics, to provide experience feedback on its operation and to present its potential. (author)

  4. An overview of fuel-coolant interactions (FCI) research at NRC

    Energy Technology Data Exchange (ETDEWEB)

    Basu, S.; Speis, T.P. [Nuclear Regulatory Commission, North Bethesda, MD (United States)

    1996-03-01

    An overview of the fuel-coolant interactions (FCI) research programs sponsored by the U.S. Nuclear Regulatory Commission (NRC) is presented in this paper. A historical perspective of the program is provided with particular reference to in-vessel steam explosion and its consequences on the reactor pressure vessel and the containment integrity. Emphasis is placed on research in the last decade involving fundamentals of FCI phenomenology, namely, premixing, triggering, propagation, and energetics. The status of the current understanding of in-vessel steam explosion-induced containment failure (alpha-mode) issue, and other FCI issues related to reactor vessel and containment integrity are reported, including the extensive review and discussion of these issues at the recently held second Steam Explosion Review Group Workshop (SERG-2). Ongoing NRC research programs are discussed in detail. Future research programs including those recommended at the SERG-2 workshop are outlined.

  5. Piston slap induced pressure fluctuation in the water coolant passage of an internal combustion engine

    Science.gov (United States)

    Ohta, Kazuhide; Wang, Xiaoyu; Saeki, Atsushi

    2016-02-01

    Liner cavitation is caused by water pressure fluctuation in the water coolant passage (WCP). When the negative pressure falls below the saturated vapor pressure, the impulsive pressure following the implosion of cavitation bubbles causes cavitation erosion of the wet cylinder liner surface. The present work establishes a numerical model for structural-acoustic coupling between the crankcase and the acoustic field in the WCP considering their dynamic characteristics. The coupling effect is evaluated through mutual interaction terms that are calculated from the mode shapes of the acoustic field and of the crankcase vibration on the boundary. Water pressure fluctuations in the WCP under the action of piston slap forces are predicted and the contributions of the uncoupled mode shapes of the crankcase and the acoustic field to the pressure waveform are analyzed. The influence of sound speed variations on the water pressure response is discussed, as well as the pressure on the thrust sides of the four cylinders.

  6. Influence of coolant temperature and pressure on destructive forces at fuel failure in the NSRR experiment

    Energy Technology Data Exchange (ETDEWEB)

    Kusagaya, Kazuyuki [Global Nuclear Fuel - Japan Co., Ltd., Yokosuka, Kanagawa (Japan); Sugiyama, Tomoyuki; Nakamura, Takehiko; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-01-01

    In order to design a new experimental capsule to be used in the NSRR (Nuclear Safety Research Reactor) experiment with the temperature and pressure conditions in a typical commercial BWR, coolant temperature and pressure influence is estimated for destructive forces during fuel rod failure in the experiment simulating reactivity-initiated accident (RIA). Considering steam property dependence on temperature and pressure, it is qualitatively shown that the destructive forces in the BWR operation condition are smaller than those in the room temperature and atmospheric pressure condition. Water column velocity, which determines impact by water hammer, is further investigated quantitatively by modeling the experimental system and water hammer phenomenon. As a result, the maximum velocity of the water column in the BWR operation conditions is calculated to be only about 10% of that in the room temperature and atmospheric pressure condition. (author)

  7. Higher dimensional loop quantum cosmology

    Science.gov (United States)

    Zhang, Xiangdong

    2016-07-01

    Loop quantum cosmology (LQC) is the symmetric sector of loop quantum gravity. In this paper, we generalize the structure of loop quantum cosmology to the theories with arbitrary spacetime dimensions. The isotropic and homogeneous cosmological model in n+1 dimensions is quantized by the loop quantization method. Interestingly, we find that the underlying quantum theories are divided into two qualitatively different sectors according to spacetime dimensions. The effective Hamiltonian and modified dynamical equations of n+1 dimensional LQC are obtained. Moreover, our results indicate that the classical big bang singularity is resolved in arbitrary spacetime dimensions by a quantum bounce. We also briefly discuss the similarities and differences between the n+1 dimensional model and the 3+1 dimensional one. Our model serves as a first example of higher dimensional loop quantum cosmology and offers the possibility to investigate quantum gravity effects in higher dimensional cosmology.

  8. Reverse osmosis for the recovery of boric acid from the primary coolant at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bártová, Šárka, E-mail: sarka.bartova@cvrez.cz [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); Kůs, Pavel [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); Skala, Martin [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); University of Chemical Technology, Prague, Department of Chemical Engineering, Technická 5, Prague 166 28 (Czech Republic); Vonková, Kateřina [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic)

    2016-04-15

    Highlights: • RO membranes tested for boric acid recovery from primary coolant of nuclear power plants. • Scanning electron microscopy was used for the characterization of the membranes. • Lab scale experiments performed under various operation conditions. • We proposed configuration of and operation conditions for RO unit in nuclear power plant. - Abstract: At nuclear power plants (NPP), evaporators are used for the treatment of primary coolant and other liquid radioactive waste containing H{sub 3}BO{sub 3}. Because the operation of evaporators is expensive, a number of more cost-effective alternatives has been considered, one of which is reverse osmosis. We tested reverse osmosis modules from several manufactures on a batch laboratory apparatus. SEM images of the tested membranes were taken to distinguish the differences between the membranes. Water permeability through membranes was evaluated from the experiments with pure water. The experiments were performed with feed solutions containing various concentrations of H{sub 3}BO{sub 3} in a range commonly occurring in radioactive waste. The pH of the feed solutions ranged from 5.2 to 11.2. Our results confirmed that the pH of the feed solution plays the most important role in membrane separation efficiency of H{sub 3}BO{sub 3}. Certain modifications to the pH of the feed solution were needed to enable the tested membranes to concentrate the H{sub 3}BO{sub 3} in the retentate stream, separate from the pure water in the permeate stream. On this basis, we propose the configuration of and operational conditions for a reverse osmosis unit at NPP.

  9. Source Term Analysis for Reactor Coolant System with Consideration of Fuel Burnup

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yu Jong; Ahn, Joon Gi; Hwang, Hae Ryong [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    The radiation source terms in reactor coolant system (RCS) of pressurized water reactor (PWR) are basic design information for ALARA design such as radiation protection and shielding. Usually engineering companies own self-developed computer codes to estimate the source terms in RCS. DAMSAM and FIPCO are the codes developed by engineering companies. KEPCO E and C has developed computer code, RadSTAR, for use in the Radiation Source Term Analysis for Reactor coolant system during normal operation. The characteristics of RadSTAR are as follows. (1) RadSTAR uses fuel inventory data calculated by ORIGEN, such as ORIGEN2 or ORIGEN-S to consider effects of the fuel burnup. (2) RadSTAR estimates fission products by using finite differential method and analytic method to minimize numerical error. (3) RadSTAR enhances flexibility by adding the function to build the nuclide data library (production pathway library) for user-defined nuclides from ORIGEN data library. (4) RadSTAR consists of two modules. RadSTAR-BL is to build the nuclide data library. RadSTAR-ST is to perform numerical analysis on source terms. This paper includes descriptions on the numerical model, the buildup of nuclide data library, and the sensitivity analysis and verification of RadSTAR. KEPCO E and C developed RadSTAR to calculate source terms in RCS during normal operation. Sensitivity analysis and accuracy verification showed that RadSTAR keeps stability at Δt of 0.1 day and gives more accurate results in comparison with DAMSAM. After development, RadSTAR will replace DAMSAM. The areas, necessary to further development of RadSTAR, are addition of source term calculations for activation products and for shutdown operation.

  10. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  11. Uranyl Nitrate Flow Loop

    Energy Technology Data Exchange (ETDEWEB)

    Ladd-Lively, Jennifer L [ORNL

    2008-10-01

    The objectives of the work discussed in this report were to: (1) develop a flow loop that would simulate the purified uranium-bearing aqueous stream exiting the solvent extraction process in a natural uranium conversion plant (NUCP); (2) develop a test plan that would simulate normal operation and disturbances that could be anticipated in an NUCP; (3) use the flow loop to test commercially available flowmeters for use as safeguards monitors; and (4) recommend a flowmeter for production-scale testing at an NUCP. There has been interest in safeguarding conversion plants because the intermediate products [uranium dioxide (UO{sub 2}), uranium tetrafluoride (UF{sub 4}), and uranium hexafluoride (UF{sub 6})] are all suitable uranium feedstocks for producing special nuclear materials. Furthermore, if safeguards are not applied virtually any nuclear weapons program can obtain these feedstocks without detection by the International Atomic Energy Agency (IAEA). Historically, IAEA had not implemented safeguards until the purified UF{sub 6} product was declared as feedstock for enrichment plants. H. A. Elayat et al. provide a basic definition of a safeguards system: 'The function of a safeguards system on a chemical conversion plant is in general terms to verify that no useful nuclear material is being diverted to use in a nuclear weapons program'. The IAEA now considers all highly purified uranium compounds as candidates for safeguarding. DOE is currently interested in 'developing instruments, tools, strategies, and methods that could be of use to the IAEA in the application of safeguards' for materials found in the front end of the nuclear fuel cycle-prior to the production of the uranium hexafluoride or oxides that have been the traditional starting point for IAEA safeguards. Several national laboratories, including Oak Ridge, Los Alamos, Lawrence Livermore, and Brookhaven, have been involved in developing tools or techniques for safeguarding conversion

  12. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    Energy Technology Data Exchange (ETDEWEB)

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  13. An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

    Science.gov (United States)

    Gamble, K. A.; Barani, T.; Pizzocri, D.; Hales, J. D.; Terrani, K. A.; Pastore, G.

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Further experiments are required to confirm these observations.

  14. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    Energy Technology Data Exchange (ETDEWEB)

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  15. Conceptual design of a thermalhydraulic loop for multiple test geometries at supercritical conditions named Supercritical Phenomena Experimental Test Apparatus (SPETA)

    Science.gov (United States)

    Adenariwo, Adepoju

    The efficiency of nuclear reactors can be improved by increasing the operating pressure of current nuclear reactors. Current CANDU-type nuclear reactors use heavy water as coolant at an outlet pressure of up to 11.5 MPa. Conceptual SuperCritical Water Reactors (SCWRs) will operate at a higher coolant outlet pressure of 25 MPa. Supercritical water technology has been used in advanced coal plants and its application proves promising to be employed in nuclear reactors. To better understand how supercritical water technology can be applied in nuclear power plants, supercritical water loops are used to study the heat transfer phenomena as it applies to CANDU-type reactors. A conceptual design of a loop known as the Supercritical Phenomena Experimental Apparatus (SPETA) has been done. This loop has been designed to fit in a 9 m by 2 m by 2.8 m enclosure that will be installed at the University of Ontario Institute of Technology Energy Research Laboratory. The loop include components to safely start up and shut down various test sections, produce a heat source to the test section, and to remove reject heat. It is expected that loop will be able to investigate the behaviour of supercritical water in various geometries including bare tubes, annulus tubes, and multi-element-type bundles. The experimental geometries are designed to match the fluid properties of Canadian SCWR fuel channel designs so that they are representative of a practical application of supercritical water technology in nuclear plants. This loop will investigate various test section orientations which are the horizontal, vertical, and inclined to investigate buoyancy effects. Frictional pressure drop effects and satisfactory methods of estimating hydraulic resistances in supercritical fluid shall also be estimated with the loop. Operating limits for SPETA have been established to be able to capture the important heat transfer phenomena at supercritical conditions. Heat balance and flow calculations have

  16. Modeling loop entropy.

    Science.gov (United States)

    Chirikjian, Gregory S

    2011-01-01

    Proteins fold from a highly disordered state into a highly ordered one. Traditionally, the folding problem has been stated as one of predicting "the" tertiary structure from sequential information. However, new evidence suggests that the ensemble of unfolded forms may not be as disordered as once believed, and that the native form of many proteins may not be described by a single conformation, but rather an ensemble of its own. Quantifying the relative disorder in the folded and unfolded ensembles as an entropy difference may therefore shed light on the folding process. One issue that clouds discussions of "entropy" is that many different kinds of entropy can be defined: entropy associated with overall translational and rotational Brownian motion, configurational entropy, vibrational entropy, conformational entropy computed in internal or Cartesian coordinates (which can even be different from each other), conformational entropy computed on a lattice, each of the above with different solvation and solvent models, thermodynamic entropy measured experimentally, etc. The focus of this work is the conformational entropy of coil/loop regions in proteins. New mathematical modeling tools for the approximation of changes in conformational entropy during transition from unfolded to folded ensembles are introduced. In particular, models for computing lower and upper bounds on entropy for polymer models of polypeptide coils both with and without end constraints are presented. The methods reviewed here include kinematics (the mathematics of rigid-body motions), classical statistical mechanics, and information theory.

  17. The loop gravity string

    CERN Document Server

    Freidel, Laurent; Pranzetti, Daniele

    2016-01-01

    In this work we study canonical gravity in finite regions for which we introduce a generalisation of the Gibbons-Hawking boundary term including the Immirzi parameter. We study the canonical formulation on a spacelike hypersuface with a boundary sphere and show how the presence of this term leads to an unprecedented type of degrees of freedom coming from the restoration of the gauge and diffeomorphism symmetry at the boundary. In the presence of a loop quantum gravity state, these boundary degrees of freedom localize along a set of punctures on the boundary sphere. We demonstrate that these degrees of freedom are effectively described by auxiliary strings with a 3-dimensional internal target space attached to each puncture. We show that the string currents represent the local frame field, that the string angular momenta represent the area flux and that the string stress tensor represents the two dimensional metric on the boundary of the region of interest. Finally, we show that the commutators of these broken...

  18. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2015-01-01

    Full Text Available The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs. Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel. The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

  19. Hard Loops, Soft Loops, and High Density Effective Field Theory

    CERN Document Server

    Schäfer, T

    2003-01-01

    We study several issues related to the use of effective field theories in QCD at large baryon density. We show that the power counting is complicated by the appearance of two scales inside loop integrals. Hard dense loops involve the large scale $mu^2$ and lead to phenomena such as screening and damping at the scale $gmu$. Soft loops only involve small scales and lead to superfluidity and non-Fermi liquid behavior at exponentially small scales. Four-fermion operators in the effective theory are suppressed by powers of $1/mu$, but they get enhanced by hard loops. As a consequence their contribution to the pairing gap is only suppressed by powers of the coupling constant, and not powers of $1/mu$. We determine the coefficients of four-fermion operators in the effective theory by matching quark-quark scattering amplitudes. Finally, we introduce a perturbative scheme for computing corrections to the gap parameter in the superfluid phase

  20. Radial propagators and Wilson loops

    CERN Document Server

    Leupold, S; Leupold, Stefan; Weigert, Heribert

    1996-01-01

    We present a relation which connects the propagator in the radial (Fock-Schwinger) gauge with a gauge invariant Wilson loop. It is closely related to the well-known field strength formula and can be used to calculate the radial gauge propagator. The result is shown to diverge in four-dimensional space even for free fields, its singular nature is however naturally explained using the renormalization properties of Wilson loops with cusps and self-intersections. Using this observation we provide a consistent regularization scheme to facilitate loop calculations. Finally we compare our results with previous approaches to derive a propagator in Fock-Schwinger gauge.

  1. CNP1000轴封式核主泵技术%Technology for CNP1000 shaft sealed reactor coolant pump

    Institute of Scientific and Technical Information of China (English)

    冯晓东; 吴大转; 杨立峰; 贾允

    2016-01-01

    CNP1000轴封式核主泵作为泵类行业的高端设备是核电站一回路系统中唯一没有实现国产化的设备,其自主设计、制造及试验是中国推进核电自主化的重点和难点。对核电泵的国内技术现状以及1000 MW轴封式核主泵结构特点进行了介绍,分别从承压边界、水力部件、转子轴系、隔热组件、轴密封、停车密封、水导轴承、推力轴承、一体化供油泵、油密封、辅助系统及其关键设备进行了设计原理和实施要求的详细说明。通过对不同类型主泵的出口限制情况和应用前景比较分析,说明了轴封主泵对外出口的优势,提出轴封主泵将在很长一段时间占据核电市场的主导地位。目前国内企业已完成了多项关键技术,如主轴密封、水力部件、全流量试验等的自主化,但是核主泵的理论基础并没有完整建立,有待进一步研究。因此,核主泵的国产化研制任务仍需较长的时间去攻克。%As high-end equipment in pump industry , CNP1000 shaft sealed reactor coolant pump ( RCP) in the primary loop system of pressurized water reactor ( PWR) in a nuclear power plant is the only equipment which cannot be manufactured domestically .Therefore, the independent design , ma-nufacturing and test of this kind of pump are important and difficult tasks in promoting the nationaliza-tion of nuclear technology in China .The domestic technological situation and structural characteristics of 1 000 MW shaft sealed RCP are introduced , and the design principle and technical implementation requirement are elaborated in terms of pressure retaining boundary , hydraulic components , pump rotor system, heat insulation subassembly , shaft seal, standstill seal, lower guide bearings, thrust bearing, integrated oil pump , oil seal, auxiliary system and key equipment in detail .Through an analysis on ex-port restriction and application prospect of various types of RCP , it is

  2. Specifics of high-temperature sodium coolant purification technology in fast reactors for hydrogen production and other innovative applications

    Directory of Open Access Journals (Sweden)

    F.A. Kozlov

    2017-03-01

    Full Text Available In creating a large-scale atomic-hydrogen power industry, the resolution of technological issues associated with high temperatures in reactor plants (900°C and large hydrogen concentrations intended as long-term resources takes on particular importance. The paper considers technological aspects of removing impurities from high-temperature sodium used as a coolant in the high-temperature fast reactor (BN-HT 600MW (th. intended for the production of hydrogen as well as other innovative applications. The authors examine the behavior of impurities in the BN-HT circuits associated with the mass transfer intensification at high temperatures (Arrhenius law in different operating modes. Special attention is given to sodium purification from hydrogen, tritium and corrosion products in the BN-HT. Sodium purification from hydrogen and tritium by their evacuation through vanadium or niobium membranes will make it possible to develop compact highly-efficient sodium purification systems. It has been shown that sodium purification from tritium to concentrations providing the maximum permissible concentration of the produced hydrogen (3.6Bq/l according to NRB-99/2009 specifies more stringent requirements to the hydrogen removal system, i.e., the permeability index of the secondary tritium removal system should exceed 140kg/s. Provided that a BN-HN-type reactor meets these conditions, the bulk of tritium (98% will be accumulated in the compact sodium purification system of the secondary circuit, 0.6% (∼ 4·104Bq/s, will be released into the environment and 1.3% will enter the product (hydrogen. The intensity of corrosion products (CPs coming into sodium is determined by the corrosion rate of structural materials: at a high temperature level, a significant amount of corrosion products flows into sodium. The performed calculations showed that, for the primary BN-HT circuit, the amount of corrosion products formed at the oxygen concentration in sodium of 1mln

  3. Product Integrals and Wilson loops

    CERN Document Server

    Karp, R L

    2001-01-01

    Using product integrals we review the unambiguous mathematical representation of Wilson line and Wilson loop operators, including their behavior under gauge transformations and the non-abelian Stokes theorem. Interesting consistency conditions among Wilson lines are also presented.

  4. Thermal fluctuations in loop cosmology

    CERN Document Server

    Magueijo, J; Magueijo, Joao; Singh, Parampreet

    2007-01-01

    Quantum gravitational effects in loop quantum cosmology lead to a resolution of the initial singularity and have the potential to solve the horizon problem and generate a quasi scale-invariant spectrum of density fluctuations. We consider loop modifications to the behavior of the inverse scale factor below a critical scale in closed models and assume a purely thermal origin for the fluctuations. We show that the no-go results for scale invariance in classical thermal models can be evaded even if we just consider modifications to the background (zeroth order) gravitational dynamics. Since a complete and systematic treatment of the perturbed Einstein equations in loop cosmology is still lacking, we simply parameterize their expected modifications. These change quantitatively, but not qualitatively, our conclusions. We thus urge the community to more fully work out this complex aspect of loop cosmology, since the full picture would not only fix the free parameters of the theory, but also provide a model for a no...

  5. Loop Quantum Cosmology Gravitational Baryogenesis

    CERN Document Server

    Odintsov, S D

    2016-01-01

    Loop Quantum Cosmology is an appealing quantum completion of classical cosmology, which brings along various theoretical features which in many cases offer remedy or modify various classical cosmology aspects. In this paper we address the gravitational baryogenesis mechanism in the context of Loop Quantum Cosmology. As we demonstrate, when Loop Quantum Cosmology effects are taken into account in the resulting Friedmann equations for a flat Friedmann-Robertson-Walker Universe, then even for a radiation dominated Universe, the predicted baryon-to-entropy ratio from the gravitational baryogenesis mechanism is non-zero, in contrast to the Einstein-Hilbert case, in which case the baryon-to-entropy ratio is zero. We also discuss various other cases apart from the radiation domination case, and we discuss how the baryon-to-entropy ratio is affected from the parameters of the quantum theory. In addition, we use illustrative exact solutions of Loop Quantum Cosmology and we investigate under which circumstances the bar...

  6. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-11-01

    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  7. Continuous smearing of Wilson Loops

    CERN Document Server

    Lohmayer, Robert

    2011-01-01

    Continuum smearing was introduced in section 4.1 of JHEP03, 064 (2006) as a meaningful continuum analogue of the well known set of lattice techniques by the same name. Here we apply continuous smearing in continuous space-time to Wilson loops in order to clarify what it does in the context of field theory and also in the context of the loop calculus of the Makeenko-Migdal equation.

  8. The Projectile inside the Loop

    OpenAIRE

    Varieschi, Gabriele U.

    2005-01-01

    In this paper we describe an alternative use of the loop-the-loop apparatus, which can be used to study an interesting case of projectile motion. We also present an effective way to perform and analyze these experiments, by using video capture software together with a digital video camera. These experiments can be integrated into classroom demonstrations for general physics courses, or become part of laboratory activities.

  9. Introduction to Loop Quantum Gravity

    OpenAIRE

    Mercuri, Simone

    2010-01-01

    The questions I have been asked during the 5th International School on Field Theory and Gravitation, have compelled me to give an account of the premises that I consider important for a beginner's approach to Loop Quantum Gravity. After a description of some general arguments and an introduction to the canonical theory of gravity, I review the background independent approach to quantum gravity, giving only a brief survey of Loop Quantum Gravity.

  10. Bifurcations of nontwisted heteroclinic loop

    Institute of Scientific and Technical Information of China (English)

    田清平; 朱德明

    2000-01-01

    Bifurcations of nontwisted and fine heteroclinic loops are studied for higher dimensional systems. The existence and its associated existing regions are given for the 1-hom orbit and the 1-per orbit, respectively, and bifurcation surfaces of the two-fold periodic orbit are also obtained. At last, these bifurcation results are applied to the fine heteroclinic loop for the planar system, which leads to some new and interesting results.

  11. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  12. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

  13. Analysis of the impact of coolant density variations in the high efficiency channel of a pressure tube super critical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scriven, M.G.; Hummel, D.W.; Novog, D.R.; Luxat, J.C. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    The Pressure Tube (PT) Supercritical Water Reactor (SCWR) is based on a light water coolant operating at pressures above the thermodynamic critical pressure; a separate low temperature and low pressure moderator. The coolant density changes by an order of magnitude depending on its local enthalpy in the porous ceramic insulator tube. This causes significant changes in the neutron transport characteristics, axially and radially, in the fuel channel. This work performs lattice physics calculations for a 78-element Pu-Th fuel at zero burnup and examines the effect of assumptions related to coolant density in the radial direction of a HEC, using the neutron transport code WIMS-AECL. (author)

  14. Bol loops of odd prime exponent

    CERN Document Server

    Foguel, Tuval

    2009-01-01

    Although any finite Bol loop of odd prime exponent is solvable, we show there exist such Bol loops with trivial center. We also construct finitely generated, infinite, simple Bruck loops of odd prime exponent for sufficiently large primes. This shows that the Burnside problem for Bruck loops has a negative answer.

  15. Classifying Finitely Generated Indecomposable RA Loops

    CERN Document Server

    Cornelissen, Mariana

    2012-01-01

    In 1995, E. Jespers, G. Leal and C. Polcino Milies classified all finite ring alternative loops (RA loops for short) which are not direct products of proper subloops. In this paper we extend this result to finitely generated RA loops and provide an explicit description of all such loops.

  16. Kalman Orbit Optimized Loop Tracking

    Science.gov (United States)

    Young, Lawrence E.; Meehan, Thomas K.

    2011-01-01

    Under certain conditions of low signal power and/or high noise, there is insufficient signal to noise ratio (SNR) to close tracking loops with individual signals on orbiting Global Navigation Satellite System (GNSS) receivers. In addition, the processing power available from flight computers is not great enough to implement a conventional ultra-tight coupling tracking loop. This work provides a method to track GNSS signals at very low SNR without the penalty of requiring very high processor throughput to calculate the loop parameters. The Kalman Orbit-Optimized Loop (KOOL) tracking approach constitutes a filter with a dynamic model and using the aggregate of information from all tracked GNSS signals to close the tracking loop for each signal. For applications where there is not a good dynamic model, such as very low orbits where atmospheric drag models may not be adequate to achieve the required accuracy, aiding from an IMU (inertial measurement unit) or other sensor will be added. The KOOL approach is based on research JPL has done to allow signal recovery from weak and scintillating signals observed during the use of GPS signals for limb sounding of the Earth s atmosphere. That approach uses the onboard PVT (position, velocity, time) solution to generate predictions for the range, range rate, and acceleration of the low-SNR signal. The low- SNR signal data are captured by a directed open loop. KOOL builds on the previous open loop tracking by including feedback and observable generation from the weak-signal channels so that the MSR receiver will continue to track and provide PVT, range, and Doppler data, even when all channels have low SNR.

  17. Assessment of ISLOCA risk: Methodology and application to a Westinghouse four-loop ice condenser plant

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.; Auflick, J.L.; Haney, L.N. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1992-04-01

    Inter-system loss-of-coolant accidents (ISLOCAs) have been identified as important contributors to offsite risk for some nuclear power plants. A methodology has been developed for identifying and evaluating plant-specific hardware designs, human factors issues, and accident consequence factors relevant to the estimation of ISLOCA core damage frequency and risk. This report presents a detailed description of the application of this analysis methodology to a Westinghouse four-loop ice condenser plant. This document also includes appendices A through I which provide: System descriptions; ISLOCA event trees; human reliability analysis; thermal hydraulic analysis; core uncovery timing calculations; calculation of system rupture probability; ISLOCA consequences analysis; uncertainty analysis; and component failure analysis.

  18. Interaction study between MOX fuel and eutectic lead-bismuth coolant

    Science.gov (United States)

    Vigier, Jean-François; Popa, Karin; Tyrpekl, Vaclav; Gardeur, Sébastien; Freis, Daniel; Somers, Joseph

    2015-12-01

    In the frame of the MYRRHA reactor project, the interaction between fuel pellets and the reactor coolant is essential for safety evaluations, e.g. in case of a pin breach. Therefore, interaction tests between uranium-plutonium mixed oxide (MOX) pellets and molten lead bismuth eutectic (LBE) have been performed and three parameters were studied, namely the interaction temperature (500 °C and 800 °C), the oxygen content in LBE and the stoichiometry of the MOX (U0.7Pu0.3O2-x and U0.7Pu0.3O2.00). After 50 h of interaction in closed containers, the pellet integrity was preserved in all cases. Whatever the conditions, neither interaction compounds (crystalline or amorphous) nor lead and bismuth diffusion into the surface regions of the MOX pellets has been detected. In most of the conditions, actinide releases into LBE were very limited (in the range of 0.01-0.15 mg), with a homogeneous release of the different actinides present in the MOX. Detected values were significantly higher in the 800 °C and low LBE oxygen content tests for both U0.7Pu0.3O2-x and U0.7Pu0.3O2.00, with 1-2 mg of actinide released in these conditions.

  19. Multiple Boundary Layer Stripping Model by Plateau-Rayleigh Instability for Fuel-Coolant Interactions

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Woo Hyun; Moriyama, Kiyofumi; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of)

    2016-05-15

    One of them is FCI (Fuel-Coolant Interaction) phenomenon which is resulted from RPV (Reactor Pressure Vessel) failure at high pressure and high temperature condition of molten fuel. If RPV fails, the melt is ejected to the cavity which is flooded by water as a jet form. Then, the ejected melt jet interacts with water causing massive steam generation and resulting in particulate debris bed on the basemat. As a result of FCI, the initial boundary conditions of steam explosion and debris bed coolability are determined and that is the reason why understanding exact mechanism of melt jet breakup is important in this field. That is, FCI can be said as a starting phenomenon in the ex-vessel severe accident scenario. Until now, numerous previous researchers conducted FCI experiments and numerical analysis in small scale and plant scale. In two MATE experiments, the jet breakup lengths are compared and analyzed with the visualization data. From the observation, the new jet breakup model is proposed including the multiple boundary layer stripping mechanism. Combining the existing and new models, the erosion rate fraction for total melt mass rate was obtained. The new model showed that multiple BLS mechanisms contribute approximately 30% of the total melt jet breakup resulting in the short jet breakup length observed in the MATE 00-2 experiment.

  20. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F.; Gauthier, G.; Carlin, F. [and others

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40{degrees}C or 70{degrees}C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased.

  1. Aging and loss-of-coolant accident (LOCA) testing of electrical connections

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01

    This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 types of cable-to-device connectors, 3 types of cable-to-cable connectors, and 2 types of in-line splices. The connections were aged for 6 months under simultaneous thermal (99 C) and radiation (46 Gy/hr) conditions. A simulated LOCA consisting of sequential high dose-rate irradiation (3 kGy/hr) and high-temperature steam exposures followed the aging. Connection functionality was monitored using insulation resistance measurements during the aging and LOCA exposures. Because only 5 of the 10 connection types passed a post-LOCA, submerged dielectric withstand test, further detailed investigation of electrical connections and the effects of cable jacket integrity on the cable-connection system is warranted.

  2. FEM Analysis and Experimental Verification of the Integral Forging Process for AP1000 Primary Coolant Pipe

    Science.gov (United States)

    Wang, Shenglong; Yu, Xiaoyi; Yang, Bin; Zhang, Mingxian; Wu, Huanchun

    2016-10-01

    AP1000 primary coolant pipes must be manufactured by integral forging technology according to the designer—Westinghouse Electric Co. The characteristics of these large, special-shaped pipes create nonuniform temperatures, effective stress, and effective strain during shaping of the pipes. This paper presents a three-dimensional finite element simulation (3D FEM) of the integral forging process, and qualitatively evaluates the likelihood of forging defects. By analyzing the evolution histories of the three field variables, we concluded that the initial forging temperature should be strictly controlled within the interval 1123 K to 1423 K (850 °C to 1150 °C) to avoid second-phase precipitation. In the hard deformation zones, small strains do not contribute to recrystallization resulting in coarse grains. Conversely, in the free deformation zone, the large strains can contribute to the dynamic recrystallization, favoring grain refinement and closure of voids. Cracks are likely to appear, however, on the workpiece surface when forging leads to large deformations. Based on the simulation results, an eligible workpiece with good mechanical properties, few macroscopic defects, and favorable grain size has been successfully forged by experiments at an industrial scale, which validates the FEM simulation.

  3. Jet model for slot film cooling with effect of free-stream and coolant turbulence

    Science.gov (United States)

    Simon, Frederick F.

    1986-01-01

    An analysis was performed utilizing the model of a wall jet for obtaining equations that will predict slot film-cooling efficiency under conditions of variable turbulence intensity, flow, and temperature. The analysis, in addition to assessing the effects of the above variables, makes a distinction between an initial region and a fully developed region. Such a distinction is important in determining the role that the turbulence intensity of the coolant plays in effecting film-cooling effectiveness in the area of the slot exit. The results of the analysis were used in the correlation of the results of a well-designed film-cooling experiment. The result of the analysis and experiment was equations that predicted film-cooling efficiency within + or - 4% average deviation for lateral free-stream turbulence intensities up to 24% and blowing rates up to 1.9. These equations should be useful in determining the optimum quantity of cooling air requried for protecting the wall of a combustor.

  4. Strategies for reactor safety: Preventing loss of coolant accidents. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B.O.Y. [RSA Technoligies, Vista (United States)

    1997-12-01

    This final report on the NKS/RAK-1.2 summarizes the main features of the PIFRAP PC-program and its intended implementation. Regardless of the preferred technical approach to LOCA frequency estimation, the analysis approach must include recognition of the following technical issues: (a) Degradation and failure mechanisms potentially affecting piping systems within the reactor coolant pressure boundary (RCPB) and the potential consequences; (b) In-service inspection practices and how they influence piping reliability; and (c) The service experience with piping systems. The report consists of six sections and one appendix. A Nordic perspective on LOCA and nuclear safety is given. It includes summaries of results from research in material sciences and current regulatory philosophies regarding piping reliability. A summary of the LOCA concept is applied in Nordic PSA studies. It includes a discussion on deterministic and probabilistic views on LOCA. The R and D on piping reliability by SKI and the PIFRAP model is summarized. Next, Section 6 presents conclusion and recommendations. Finally, Appendix A contains a list of abbreviations and acronyms, together with a glossary of technical terms. (EG) 16 refs.

  5. Prevention of unacceptable material fatigue considering the coolant. Principles and application

    Energy Technology Data Exchange (ETDEWEB)

    Kraetschmer, Daniel; Herter, Karl-Heinz; Schuler, Xaver [Stuttgart Univ. (Germany). MPA

    2010-07-01

    For the construction, design and operation of nuclear systems, structures and components the appropriate technical codes and standards provide material data, detailed stress analysis procedures and a design philosophy which guarantees a reliable behavior of the structural components throughout the specified lifetime. Especially for cyclic stress evaluation the different codes and standards provide different fatigue analyses procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. For the fatigue design curves used as limiting criteria the influence of different factors like e.g., environment, surface finish and temperature must be taken into consideration in an appropriate way. A general numerical calculation procedure was developed to calculate equivalent stress- and strain ranges, according to different technical codes and standards. The additional implementation of already published environmental correction factors, depending on actual temperature, strain rate and dissolved-oxygen level, allows the practical and fast application of a strain-based approach to evaluate fatigue at varying temperatures and strain rates for specimens and components exposed to coolant environment. Proposed new design curves with and without the incorporation of environmental effects as well as design curves according to current technical codes and standards are considered. The application of this procedure is demonstrated and discussed by the example of a pressurizer nozzle under transient stratification loads, measured by in-service monitoring. (orig.)

  6. Thermal hydraulic investigations of primary coolant pipe rupture in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)]. E-mail: natesan@igcar.ernet.in; Kasinathan, N. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Velusamy, K. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Selvaraj, P. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Chellapandi, P. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Chetal, S.C. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2006-06-15

    High quality for primary coolant pipes in fast reactors is ensured through utmost care taken in the design and manufacture. Demonstration of high structural reliability of them by extensive experimental and theoretical studies renders the double-ended guillotine rupture (DEGR) of a primary pipe a highly improbable event. However, as a defense in depth approach instantaneous DEGR of one of the pipes has been considered in design. Thermal hydraulic analyses of this event in a typical liquid metal cooled fast breeder have been carried out to study its consequences and to establish the availability of safety margins. Various uncertainties relevant to the event have been analysed to evaluate the sensitivity of each parameter. For this purpose, one-dimensional plant dynamics studies using thermal and hydraulic models of core subassemblies and primary sodium circuit have been performed. Validity of the assumptions made in the one-dimensional model like, uniform flow through all subassemblies in core under pipe ruptured condition and non possibility of sodium boiling by flashing have also been investigated through detailed three-dimensional and pressure transient studies. Analyses indicate the availability of good margins against the design safety limits in all the parametric cases analysed.

  7. Liquid-lithium nitrate: candidate fusion reactor coolant or chemical curiosity

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M.G.

    1986-01-01

    On the basis of the recent Blanket Comparison and Selection Study, 700 K was selected as the threshold for useful high-temperature operation of a liquid nitrate coolant and 800 K was assumed as a realistic upper operating limit. Both standard Second Law equilibrium calculations and computerized Gibbs energy minimization methods were used to determine equilibrium compositions of multiphase, multicomponent salt systems at specified temperatures under certain condensed were performed on the LiNO/sub 3/-LiNO/sub 2/, NaNO/sub 3/- NaNO/sub 2/, and KNO/sub 3/-KNO/sub 2/ systems, and then predicted decomposition pressures were compared for equivalent degrees of decomposition at temperatures ranging from 600 to 900K. Two approaches were taken in calculating decomposition pressures over MNO/sub 3/-MNO/sub 2/ systems: (a) allowing the formation of molecular N/sub 2/ as a gaseous reaction product and (b) not allowing its formation. In calculations of MNO/sub 2/-M/sub 2/O-MOH-H/sub 2/O equilibria, which were used to evaluate the reversibility of tritium dissolution and release, the activity of hydroxide reaction product was determined as a function of water activity at two representative temperatures. Preliminary results and conclusions are summarized.

  8. Chemical and spectroscopic characterization of a vegetable oil used as dielectric coolant in distribution transformers

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, Neffer A.; Abonia, Rodrigo, E-mail: rodrigo.abonia@correounivalle.edu.co [Departamento de Quimica, Escuela de Ingenieria Electrica, Universidad del Valle, Cali (Colombia); Cadavid, Hector [Grupo GRALTA, Escuela de Ingenieria Electrica, Universidad del Valle, Cali (Colombia); Vargas, Ines H. [Area de Ingenieria de Distribucion, Empresas Publicas de Medellin (EPM), Medellin (Colombia)

    2011-09-15

    In this work, a complete UV-Vis, IR and (1H, 13C and DEPT) NMR spectroscopic analysis was performed for a FR3 vegetable oil sample used as dielectric coolant in an experimental distribution transformer. The same spectroscopic analysis was performed for three used FR3 oil samples (i.e., 4 months in use, 8 months in use and 7 years in use), removed from several operating distribution transformers. Comparison of the data indicated that no significant spectroscopic changes, and hence, no structural changes occurred to the oils by the use. Chemical transformations like catalytic hydrogenation (hardening) and hydrolysis were performed to the FR3 oil sample and the obtained products were analyzed by spectroscopic methods in order to collect further structural information about the FR3 oil. Accelerated aging tests in laboratory were also performed for three FR3 oil samples affording interesting information about the structure of the degradation products. These findings would be valuable to search for a spectroscopy-based technique for monitoring the lifetime and performance of this insulating vegetable oil. (author)

  9. Effects of coatings on storability of carrot under evaporative coolant system.

    Directory of Open Access Journals (Sweden)

    Adetunji Charles Oluwaseun

    2013-09-01

    Full Text Available Four different coatings were developed from the mucilage of Cactus and their effects were investigated on the quality and storability of carrot fruits. The four experimental coatings were: Pure mucilage extracts (ME, Mucilage extract mixed with 5ml glycerol (MEG, Mucilage extract mixed 5ml soy oil ( MESO, Mucilage extract mixed with 5ml olive oil(MEOO the addition of oil served as plasticizer. The following parameters were measured: weight loss, ascorbic acid content, pH, firmness and microbial qualities. Four hundred and eighty (480 carrot were arranged randomly into five treatments, the control (untreated and four coating treatments were stored for seven weeks under Evaporative Coolant System (ECS. Prior to storage, the carrot samples were surface sterilized using 100mg/L sodium hypochlorite. Results showed that Cactus mucilage was effective in extending the shelf-life of carrotwhen compared to untreated control in the following order: MESO>MEOO>MEG>ME>Control. Results revealed that coatings hindered the growth of microorganisms significantly

  10. Corrosion of high temperature alloys in the coolant helium of a gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cabet, C.; Terlain, A. [Service de la Corrosion et du Comportement des Materiaux dans leur Environnement, DEN/DPC - CEA/Saclay, Gif sur Yvette (France); Monnier, A. [Lab. de Genie Electrique de Paris, Plateau du Moulon, Gif sur Yvette (France)

    2004-07-01

    The corrosion of structural alloys in gas cooled reactor environment appears to be a critical issue. The coolant helium proved to contain impurities mainly H{sub 2}, H{sub 2}O, CO, and CH{sub 4} in the microbar range that interact with metallic materials at high temperature. Surface scale formation, bulk carburisation and/or decarburisation can occur, depending on the gas chemistry, the alloy composition and the temperature. These structural transformations can notably influence the component mechanical properties. A short review of the literature on the topic is first given. Corrosion tests with high chromium alloys and a Mo-based alloy were carried out at 750 C in a purposely-designed facility under simulated GCR helium. The first, rather short term, results showed that the Mo-based alloy was inert while the others alloys oxidised during at least 900 hours. The alloy with the higher Al and Ti contents exhibited poor oxidation resistance impeding its use as structural material without further investigations. (orig.)

  11. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  12. Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).

  13. Analysis of an AP600 intermediate-size loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Lime, J.F. [Los Alamos National Lab., NM (United States)

    1995-09-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  14. Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

    Science.gov (United States)

    Huang, J. Y.; Chiang, M. F.; Jeng, S. L.; Huang, J. S.; Kuo, R. C.

    2013-01-01

    The environmentally assisted cracking behavior of dissimilar metal (DM) welds, including Alloy 52-A 508 and Alloy 82-A508, under simulated BWR coolant conditions was studied. Effects of postweld heat treatment and sulfur content of the base metal on the corrosion fatigue and SCC growth rates of DM welds were evaluated. The crack growth rates for the DM weld heat-treated at 621 °C for 24 h were observed to be faster than those for the as-welded. But the DM weld heat-treated at 621 °C for 8 h + 400 °C for 200 h showed better SCC resistance than the as-welded. The longer the heat treatment at 621 °C, the higher the chromium carbides density along the grain boundary was observed. Sulfur could diffuse out of the base metal and segregate along the grain boundaries of the dilution zone, leading to weakening the grain boundary strength and the SCC resistance of the Alloy 52-A508 weld.

  15. A probability model: Tritium release into the coolant of a light water tritium production reactor

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D N

    1992-04-01

    This report presents a probability model of the total amount of tritium that will be released from a core of tritium target rods into the coolant of a light water reactor during a tritium production cycle.The model relates the total tritium released from a core to the release characteristics of an individual target rod within the core. The model captures total tritium release from two sources-release via target rod breach and release via permeation through the target rod. Specifically, under conservative assumptions about the breach characteristics of a target rod, total tritium released from a core is modeled as a function of the probability of a target breach and the mean and standard deviation of the permeation reduction factor (PRF) of an individual target rod. Two dominant facts emerge from the analysis in this report. First, total tritium release cannot be controlled and minimized solely through the PRF characteristics of a target rod. Tritium release via breach must be abated if acceptable tritium production is to be achieved. Second, PRF values have a saturation point to their effectiveness. Specifically, in the presence of any realistic level of PRF variability, increasing PRF values above approximately 1000 wig contribute little to minimizing total tritium release.

  16. Proceedings of the OECD/CSNI specialists meeting on fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru; Yamano, Norihiro; Sugimoto, Jun [eds.

    1998-01-01

    The OECD/CSNI Specialists Meeting on Fuel Coolant Interactions (FCI) was held at Tokai-mura in Japan on May 19 through 21, 1997, and attended by 80 participants from 14 countries and one international organizations. In the meeting 36 papers were presented followed by active discussions in six sessions on various aspects of FCI issues, such as reactor application, premixing, propagation/trigger, experiments and code/models. At the end of the Meeting, the participants have reached to the consensus on the summary and recommendations, which consists of the following items; (1) We find no new evidence that would change or violate the conclusion of SERG-2 (1996) that alpha-mode failure is not risk significant. (2) Significant progress has been made since the Santa Barbara meeting (1993). (3) Several areas have been identified, which need further investigations to understand the basic FCI phenomena, and to improve the modeling. (4) We recommend maximizing open communication between various research groups in order to accelerate the resolution of the remaining issues. (5) We recommend that the next specialist meeting be held within 3 to 5 years in order to synthesize the activities described above. (J.P.N.)

  17. Corrosion fatigue studies on F82H mod. martensitic steel in reducing water coolant environments

    Energy Technology Data Exchange (ETDEWEB)

    Maday, M.F.; Masci, A. [ENEA, Casaccia (Italy). Centro Ricerche Energia

    1998-03-01

    Load-controlled low cycle fatigue tests have been carried out on F82H martensitic steel in 240degC oxygen-free water with and without dissolved hydrogen, in order to simulate realistic coolant boundary conditions to be approached in DEMO. It was found that water independently of its hydrogen content, determined the same fatigue life reduction compared to the base-line air results. Water cracks exhibited in their first propagation stages similar fracture morphologies which were completely missing on the air cracks, and were attributed to the action of an environment related component. Lowering frequency gave rise to an increase in F82H fatigue lifetimes without any change in cracking mode in air, and to fatigue life reduction by microvoid coalescence alone in water. The data were discussed in terms of (i) frequency dependent concurrent processes for crack initiation and (ii) frequency-dependent competitive mechanisms for crack propagation induced by cathodic hydrogen from F82H corrosion. (author)

  18. Analysis of Pressure Pulsation Induced by Rotor-Stator Interaction in Nuclear Reactor Coolant Pump

    Directory of Open Access Journals (Sweden)

    Xu Zhang

    2017-01-01

    Full Text Available The internal flow of reactor coolant pump (RCP is much more complex than the flow of a general mixed-flow pump due to high temperature, high pressure, and large flow rate. The pressure pulsation that is induced by rotor-stator interaction (RSI has significant effects on the performance of pump; therefore, it is necessary to figure out the distribution and propagation characteristics of pressure pulsation in the pump. The study uses CFD method to calculate the behavior of the flow. Results show that the amplitudes of pressure pulsation get the maximum between the rotor and stator, and the dissipation rate of pressure pulsation in impellers passage is larger than that in guide vanes passage. The behavior is associated with the frequency of pressure wave in different regions. The flow rate distribution is influenced by the operating conditions. The study finds that, at nominal flow, the flow rate distribution in guide vanes is relatively uniform and the pressure pulsation amplitude is the smallest. Besides, the vortex shedding or backflow from the impeller blade exit has the same frequency as pressure pulsation but there are phase differences, and it has been confirmed that the absolute value of phase differences reflects the vorticity intensity.

  19. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    Science.gov (United States)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  20. Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B. [RSA Technologies, Vista, CA (United States)

    1999-05-01

    This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research 41 refs, figs, tabs

  1. An Analysis of an Automatic Coolant Bypass in the International Space Station Node 2 Internal Active Thermal Control System

    Science.gov (United States)

    Clanton, Stephen E.; Holt, James M.; Turner, Larry D. (Technical Monitor)

    2001-01-01

    A challenging part of International Space Station (ISS) thermal control design is the ability to incorporate design changes into an integrated system without negatively impacting performance. The challenge presents itself in that the typical ISS Internal Active Thermal Control System (IATCS) consists of an integrated hardware/software system that provides active coolant resources to a variety of users. Software algorithms control the IATCS to specific temperatures, flow rates, and pressure differentials in order to meet the user-defined requirements. What may seem to be small design changes imposed on the system may in fact result in system instability or the temporary inability to meet user requirements. The purpose of this paper is to provide a brief description of the solution process and analyses used to implement one such design change that required the incorporation of an automatic coolant bypass in the ISS Node 2 element.

  2. Effect of complicated coolant flow behavior in the ABWR lower plenum on flow distribution to the core

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Shun, E-mail: swatanabe@edu.esys.tsukuba.ac.jp [University of Tsukuba, 1-1-1 Tennoudai, Tsukuba 305-8573, Ibaraki (Japan); Abe, Yutaka, E-mail: abe@kz.tsukuba.ac.jp [University of Tsukuba, 1-1-1 Tennoudai, Tsukuba 305-8573, Ibaraki (Japan); Kaneko, Akiko, E-mail: kaneko@kz.tsukuba.ac.jp [University of Tsukuba, 1-1-1 Tennoudai, Tsukuba 305-8573, Ibaraki (Japan); Watanabe, Fumitoshi, E-mail: Watanabe.Fumitoshi@tepco.co.jp [Tokyo Electric Power Company, 4-1, Egasaki, Tsurumi, Yokohama 230-8510, Kanagawa (Japan); Tezuka, Kenichi, E-mail: tezuka.kenichi@tepco.co.jp [Tokyo Electric Power Company, 4-1, Egasaki, Tsurumi, Yokohama 230-8510, Kanagawa (Japan); Hirao, Kazuki, E-mail: hirao.kazuki@tepco.co.jp [Tokyo Electric Power Company, 4-1, Egasaki, Tsurumi, Yokohama 230-8510, Kanagawa (Japan)

    2015-12-15

    Highlights: • A 1/10th model of the lower plenum was constructed to obtain the benchmark data. • Complicated flow behaviors were measured using LDV and PIV. • Flow distribution in the lower plenum became non-uniform. • Flow distribution became uniform at the core inlet through the side entry orifices. • The vortices were observed around the side entry orifices. - Abstract: One of the cost reduction strategies in nuclear power generation is the augmentation of power outputs by increasing the coolant flow rate. To achieve augmentation of the power outputs in an Advanced Boiling Water Reactor, it is extremely important to evaluate the coolant flow from the lower plenum to the core inlet, which dominates the coolant flow distribution over the rod bundle of the core. In the lower plenum, there are a number of structures, such as control rod driving mechanisms and guide tubes. In addition, core support beams and side entry orifices are placed at the core inlet. Therefore, it is estimated that the coolant flow becomes very complicated in these areas. To predict complicated coolant flow in the lower plenum and core inlet, simulation using a Computational Fluid Dynamics (CFD) code is helpful. However, previous experimental data are not enough to verify the simulation results for the intended flow channel in this study. Hence, this study focuses on establishing a benchmark for the CFD code using the visualization method. Based on the validation process, the functions of the complicated flow structure at the core inlet and through the lower plenum are demonstrated in detail in the coolant flow distribution to the core. A 1/10th model of a lower plenum was constructed, and the velocity profiles were measured using Particle Image Velocimetry (PIV) and Laser-Doppler Velocimetry (LDV). Each measurement was performed for a Reynolds number of 2620. From the flow velocity measurement, the vertical velocity at the center of the lower plenum is determined to be faster than

  3. Development of seismic sloshing analysis method of liquid coolant sodium in the KALIMER reactor vessel including several cylindrical components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Yoo, Bong

    2000-11-01

    It is important to establish a highly accurate technique of evaluating the sloshing behavior of liquid sodium coolant during earthquake for structural integrity of KALIMER reactor vessel and internals. The analysis procedure of sloshing behaviors is established using finite element computer program ANSYS, and the effectiveness of the procedure is confirmed by comparison with theoretical and experimental results in the literature. The analysis results agree well with experimental ones. Based on the procedure, the sloshing characteristics of liquid sodium coolant in the KALIMER reactor vessel including reactor internal components are evaluated. The maximum response height of sodium free surface at the reactor vessel is about 55cm when subjected to horizontal safe shutdown earthquake (SSE) of 0.3g for seismically isolated reactor building.

  4. Study of the Open Loop and Closed Loop Oscillator Techniques

    Energy Technology Data Exchange (ETDEWEB)

    Imel, George R. [Idaho State Univ., Pocatello, ID (United States); Baker, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riley, Tony [Knolls Atomic Power Lab. (KAPL), Schenectady, NY (United States); Langbehn, Adam [Puget Sound Naval Base, Bremerton, WA (United States); Aryal, Harishchandra [Idaho State Univ., Pocatello, ID (United States); Benzerga, M. Lamine [Idaho State Univ., Pocatello, ID (United States)

    2015-04-11

    This report presents the progress and completion of a five-year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques.The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this report we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign to measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems.

  5. BPS Wilson Loops on S^2 at Higher Loops

    CERN Document Server

    Young, Donovan

    2008-01-01

    We consider supersymmetric Wilson loops of the variety constructed by Drukker, Giombi, Ricci, and Trancanelli, whose spatial contours lie on a two-sphere. Working to second order in the 't Hooft coupling in planar N=4 Supersymmetric Yang-Mills Theory (SYM), we compute the vacuum expectation value of a wavy-latitude and of a loop composed of two longitudes. We evaluate the resulting integrals numerically and find that the results are consistent with the zero-instanton sector calculation of Wilson loops in 2-d Yang-Mills on S^2 performed by Bassetto and Griguolo. We also consider the connected correlator of two distinct latitudes to third order in the 't Hooft coupling in planar N=4 SYM. We compare the result in the limit where the latitudes become coincident to a perturbative calculation in 2-d Yang-Mills on S^2 using a light-cone Wu-Mandelstam-Leibbrandt prescription. The two calculations produce differing results.

  6. A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

    Energy Technology Data Exchange (ETDEWEB)

    Lister, D. [University of New Brunswick, Fredericton, NB (Canada). Dept. of Chemical Engineering; Lang, L.C. [Atomic Energy of Canada Ltd., Chalk River Lab., ON (Canada)

    2002-07-01

    Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

  7. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    OpenAIRE

    Hutli Ezddin; Gottlasz Valer; Tar Dániel; Ezsol Gyorgy; Baranyai Gabor

    2015-01-01

    The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs). Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image V...

  8. Performance Investigation of Automobile Radiator Operated with ZnFe2O4 Nano Fluid based Coolant

    Directory of Open Access Journals (Sweden)

    Tripathi Ajay

    2015-01-01

    Full Text Available The cooling system of an Automobile plays an important role in its performance, consists of two main parts, known as radiator and fan. Improving thermal efficiency of engine leads to increase the engine's performance, decline the fuel consumption and decrease the pollution emissions. Water and ethylene glycol as conventional coolants have been widely used in radiators of an automotive industry for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, “nanofluids” have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the preparation of Zinc based nanofluids (ZnFe2O4 using chemical co-precipitation method and its application in an automotive cooling system along with mixture of ethylene glycol and water (50:50. Relevant input data, nanofluids properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nano fluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the base-fluid compared to ethylene glycol (i.e. base-fluid alone. It is observed that, about 78% of heat transfer enhancement could be achieved with the addition of 1% ZnFe2O4 particles in a base fluid at the Reynolds number of 84.4x103 and 39.5x103 for air and coolant respectively

  9. Generalized loop space and TMDs

    Directory of Open Access Journals (Sweden)

    Mertens Tom

    2014-06-01

    Full Text Available The Standard Model describes the three (of four basic interactions known in Nature in terms of the quantum fields which are constituted by representations of special unitary gauge groups of symmetry. However, the physical observables do not always coincide with the fundamental degrees of freedom of the Standard Model. Therefore it can be useful to switch to the loop space representation of the gauge theory, where the variables are inherently gauge invariant but the degrees of freedom are absorbed in the path/loop dependence. Over-completeness of this space requires the introduction of an equivalence relation which is provided by Wilson loop functionals operating on piecewise regular paths. It is well known that certain Wilson loops show the same singularity structure as some Transverse Momentum Dependent PDFs (TMDs, which are not renormalizable by the common methods due to exactly this singularity structure. By introducing geometrical operators, like the area-derivative, we were able to derive an evolution equation for these Wilson loops and we hope to apply this method in the future to find some renormalization schemes for TMDs.

  10. Loop coupled resonator optical waveguides.

    Science.gov (United States)

    Song, Junfeng; Luo, Lian-Wee; Luo, Xianshu; Zhou, Haifeng; Tu, Xiaoguang; Jia, Lianxi; Fang, Qing; Lo, Guo-Qiang

    2014-10-06

    We propose a novel coupled resonator optical waveguide (CROW) structure that is made up of a waveguide loop. We theoretically investigate the forbidden band and conduction band conditions in an infinite periodic lattice. We also discuss the reflection- and transmission- spectra, group delay in finite periodic structures. Light has a larger group delay at the band edge in a periodic structure. The flat band pass filter and flat-top group delay can be realized in a non-periodic structure. Scattering matrix method is used to calculate the effects of waveguide loss on the optical characteristics of these structures. We also introduce a tunable coupling loop waveguide to compensate for the fabrication variations since the coupling coefficient of the directional coupler in the loop waveguide is a critical factor in determining the characteristics of a loop CROW. The loop CROW structure is suitable for a wide range of applications such as band pass filters, high Q microcavity, and optical buffers and so on.

  11. A Preliminary Study of Banana Stem Juice as a Plant-Based Coagulant for Treatment of Spent Coolant Wastewater

    Directory of Open Access Journals (Sweden)

    Habsah Alwi

    2013-01-01

    Full Text Available The effectiveness of banana stem juice as a natural coagulant for treatment of spent coolant wastewater was investigated . Three main parameters were studied, namely, chemical oxygen demand (COD, suspended solids (SSs, and turbidity of effluent. Coagulation experiments using jar test were performed with a flocculation system where the effects of spent coolant wastewater pH as well as banana stem juice dosage on coagulation effectiveness were examined. The highest recorded COD, SS, and turbidity removal percentages by banana stem juice were 80.1%, 88.6%, and 98.5%, respectively, observed for effluent at pH 7 using 90 mL dosage. The inulin concentration in the banana stem was examined to be 1.22016 mg/mL. It could be concluded that banana stem juice showed tremendous potential as a natural coagulant for water treatment purposes and could be applied in the pretreatment stage of Malaysian spent coolant wastewater prior to secondary treatment.

  12. Characterization of Industrial Coolant Fluids and Continuous Ageing Monitoring by Wireless Node-Enabled Fiber Optic Sensors.

    Science.gov (United States)

    Sachat, Alexandros El; Meristoudi, Anastasia; Markos, Christos; Sakellariou, Andreas; Papadopoulos, Aggelos; Katsikas, Serafim; Riziotis, Christos

    2017-03-11

    Environmentally robust chemical sensors for monitoring industrial processes or infrastructures are lately becoming important devices in industry. Low complexity and wireless enabled characteristics can offer the required flexibility for sensor deployment in adaptable sensing networks for continuous monitoring and management of industrial assets. Here are presented the design, development and operation of a class of low cost photonic sensors for monitoring the ageing process and the operational characteristics of coolant fluids used in an industrial heavy machinery infrastructure. The chemical, physical and spectroscopic characteristics of specific industrial-grade coolant fluids were analyzed along their entire life cycle range, and proper parameters for their efficient monitoring were identified. Based on multimode polymer or silica optical fibers, wide range (3-11) pH sensors were developed by employing sol-gel derived pH sensitive coatings. The performances of the developed sensors were characterized and compared, towards their coolants' ageing monitoring capability, proving their efficiency in such a demanding application scenario and harsh industrial environment. The operating characteristics of this type of sensors allowed their integration in an autonomous wireless sensing node, thus enabling the future use of the demonstrated platform in wireless sensor networks for a variety of industrial and environmental monitoring applications.

  13. Experimental studies into the dependences of the axial lead coolant pump performance on the impeller cascade parameters

    Directory of Open Access Journals (Sweden)

    A.V. Beznosov

    2017-06-01

    Full Text Available The paper presents results of experimental studies into the dependences of the axial lead coolant pump performance (delivery, head, efficiency on the impeller cascade parameters, including the number of blades, the cascade blade angle and the cascade solidity. The studies were conducted as applied to conditions of small and medium sized plants based on lead cooled fast neutron reactors with horizontal steam generators. The designs of such plants are now in the process of elaboration at Nizhny Novgorod State Technical University (NNSTU. The studies were conducted at NNSTU's FT-4 test facility at a lead coolant temperature of 440–500°C. In the process of investigations, the number of blades in the form of flat plates was 3, 4, 6 and 8, the cascade blade angle was in a range of 9–43°, and the cascade solidity (0.6–1.2 was changed by changing the blade section chord length. The shaft speed of the NNSTU's NSO-01 pump, onto which changeable impellers were installed, was changed in steps of 100 rev/min in an interval of 600–1100 rev/min. The blade diameter was about 200mm, and the maximum lead coolant flow rate in the course of the tests reached ∼2000t/h. The performance of 27 impellers was investigated. It is recommended that the investigation results should be used in design of axial HLMC pumps.

  14. Toyota's heat management system - coolant heat storage for mass production today, new technologies for the future

    Energy Technology Data Exchange (ETDEWEB)

    Ichinose, Hiroki; Takaoka, Toshifumi; Kobayashi, Hideo [Toyota Motor Corporation (Japan)

    2004-07-01

    There has been pressing needs for the protection of metropolitan environment and the challenge of global warming. A heat management system prevails to meet such requirements. In actual driving condition, only about 30% of the total fuel energy is consumed for propulsion and air conditioner. At the same time 60% of fuel energy is wasted as exhaust gas, thermal loss and warm up loss. It is important to manage total thermal energy as a whole vehicle to improve thermal efficiency. The principle is to reduce heat loss in order to increase exhaust gas temperature and recover heat energy for pre-heating at the next cold start. Further developed versions may include thermal management strategies including turbocharger and thermal exchanger using exhaust gas heat energy. Toyota has developed the Coolant Heat Storage system (CHSS) for one of heat management systems to reduce cold emission and improve cabin comfort. The system enables to store hot coolant at the warmed up condition in a heat storage tank. At the next cold start, it is possible for CHSS to reduce unburned hydrocarbon by preheating intake port quickly with the hot coolant in the tank. CHSS was adopted in hybrid vehicle Prius for the US model in 2003. This vehicle achieved to meet the ATPZEV,the most stringent emission regulation in the US. This paper describes a total heat management focusing on the newly developed CHSS.

  15. Vertically Polarized Omnidirectional Printed Slot Loop AntennaPrinted Slot Loop Antenna (invited)

    DEFF Research Database (Denmark)

    Kammersgaard, Nikolaj Peter Iversen; Kvist, Søren Helstrup; Thaysen, Jesper

    2015-01-01

    A novel verticall A novel vertically polarized dpolarize , omnidirection omnidirectional l , printed slot loop antenna h sprinted slot loop antenna has been designed, simulated, fabricated, and measured. The slot loop works as a magnetic loop. The loop is loaded with inductors to insure uniform...

  16. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

  17. The BLOW-3A: A theoretical model to describe transient two phase flow conditions in Liquid Metal Fast Breeder Reactor (LMFBR) coolant channels

    Science.gov (United States)

    Bottoni, M.; Struwe, D.

    The theoretical background of the BLOW-3A program is reported, including the basic equations used to determine temperature fields in the fuel, clad, coolant and structure material as well as the coolant dynamics in single and two-phase flow conditions. The two-phase flow model assumes an annular flow regime. Special aspects to calculate two-phase pressure drops for these conditions are discussed. Examples of the experimental validation of the program are given.

  18. All digital pulsewidth control loop

    Science.gov (United States)

    Huang, Hong-Yi; Jan, Shiun-Dian; Pu, Ruei-Iun

    2013-03-01

    This work presents an all-digital pulsewidth control loop (ADPWCL). The proposed system accepts a wide range of input duty cycles and performs a fast correction to the target output pulsewidth. An all-digital delay-locked loop (DLL) with fast locking time using a simplified time to digital converter and a new differential two-step delay element is proposed. The area of the delay element is much smaller than that in conventional designs, while having the same delay range. A test chip is verified in a 0.18-µm CMOS process. The measured duty cycle ranges from 4% to 98% with 7-bit resolution.

  19. Loop quantum cosmology: Recent progress

    Indian Academy of Sciences (India)

    Martin Bojowald

    2004-10-01

    Aspects of the full theory of loop quantum gravity can be studied in a simpler context by reducing to symmetric models like cosmological ones. This leads to several applications where loop effects play a significant role when one is sensitive to the quantum regime. As a consequence, the structure of and the approach to classical singularities are very different from general relativity. The quantum theory is free of singularities, and there are new phenomenological scenarios for the evolution of the very early universe such as inflation. We give an overview of the main effects, focussing on recent results obtained by different groups.

  20. Loop quantum geometry: a primer

    Energy Technology Data Exchange (ETDEWEB)

    Corichi, Alejandro [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, A. Postal 70-543, Mexico D.F. 04510 (Mexico)

    2005-01-15

    This is the written version of a lecture given at the 'VI Mexican School of Gravitation and Mathematical Physics' (Nov 21-27, 2004, Playa del Carmen, Mexico), introducing the basics of Loop Quantum Geometry. The purpose of the written contribution is to provide a Primer version, that is, a first entry into Loop Quantum Gravity and to present at the same time a friendly guide to the existing pedagogical literature on the subject. This account is geared towards graduate students and non-experts interested in learning the basics of the subject.

  1. LISA Pathfinder: OPD loop characterisation

    Science.gov (United States)

    Born, Michael; LPF Collaboration

    2017-05-01

    The optical metrology system (OMS) of the LISA Pathfinder mission is measuring the distance between two free-floating test masses with unprecedented precision. One of the four OMS heterodyne interferometers reads out the phase difference between the reference and the measurement laser beam. This phase from the reference interferometer is common to all other longitudinal interferometer read outs and therefore subtracted. In addition, the phase is fed back via the digital optical pathlength difference (OPD) control loop to keep it close to zero. Here, we analyse the loop parameters and compare them to on-ground measurement results.

  2. Loop Quantum Geometry: A primer

    OpenAIRE

    Corichi, Alejandro

    2005-01-01

    This is the written version of a lecture given at the ``VI Mexican School of Gravitation and Mathematical Physics" (Nov 21-27, 2004, Playa del Carmen, Mexico), introducing the basics of Loop Quantum Geometry. The purpose of the written contribution is to provide a Primer version, that is, a first entry into Loop Quantum Gravity and to present at the same time a friendly guide to the existing pedagogical literature on the subject. This account is geared towards graduate students and non-expert...

  3. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain); Sedano, L. A. [Asociacion Euratom-Ciematpara Fusion, Av. Complutense 22, 28040 Madrid (Spain); Perlado, J. M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain)

    2008-07-15

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T{sub 2} permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H{sub 2} tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations

  4. Effect of spray system on fission product distribution in containment during a severe accident in a two-loop pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dehjourian, Mehdi; Rahgoshay, Mohammad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-08-15

    The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  5. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

    Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  6. Selection of sodium coolant for fast reactors in the US, France and Japan

    Energy Technology Data Exchange (ETDEWEB)

    Sakamoto, Yoshihiko, E-mail: sakamoto.yoshihiko@jaea.go.jp [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan); Garnier, Jean-Claude; Rouault, Jacques [CEA, DEN, DER, Centre de Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Grandy, Christopher; Fanning, Thomas; Hill, Robert [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Chikazawa, Yoshitaka; Kotake, Shoji [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Trilateral study was conducted on coolant selection of fast reactor concept. Black-Right-Pointing-Pointer Fast reactor concepts are vital for nuclear fuel cycle sustainability goals. Black-Right-Pointing-Pointer Sodium, gas and lead cooled fast reactors are capable to achieve the goals. Black-Right-Pointing-Pointer Sodium cooled fast reactor is the most matured technology. Black-Right-Pointing-Pointer Gas and lead cooled fast reactor require long term development. - Abstract: The joint paper presents a common view of fast reactor specific missions in the development of nuclear energy and a cross-analysis of merits and demerits of several Fast Reactors concepts studied worldwide and especially in the Generation-IV International Forum (GIF) framework. The paper provides the context for fast reactors development in the United States, France and Japan and focuses on the comparison on Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), and Lead-cooled Fast Reactor (LFR), i.e. the three fast reactor concepts that have the potential to meet the nuclear fuel cycle sustainability goals. The information provided in the article permits the reader to understand each country's objectives to see that not only the objectives searched for but also the technical orientations are converging. The authors underline that SFR technology evaluation relies significantly on the substantial base technology development programs within each country which is without comparison for the other two fast reactor technologies, e.g., SFR technology has already been developed to commercial or near commercial scale in each country whereas the performance of LFR and GFR technology is still uncertain. The main GFR merits are the potential for high temperatures and the easier possibilities for inspections and repairs. The main challenges are the fuel (fabrication, in-pile behavior), materials for high temperatures, and the implementation of

  7. Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

    Science.gov (United States)

    Seguin, Nicolas

    2014-05-01

    In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical

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

    Directory of Open Access Journals (Sweden)

    Tala Moussa

    2016-01-01

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

  9. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree.

  10. Development of motors and drives for main coolant pump and CEDM

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Do Hyun; Ha, Hoi Doo; Park, Jung Woo; Koo, Dae Hyun; Chang, Ki Chan; Kim, Jong Moo; Kim, Won Ho; Rim, Geun Hie; Baek, Ju Won; Park, Doh Young; Hwang, Don Ha; Jeon, Jeong Woo [Korea Electrotechnology Research Institute, Changwon (Korea)

    1999-03-01

    A canned type 170kW induction motor for the main coolant pump (MCP) of the integral reactor SMART was designed to minimize the eddy current loss in the can and the volume of motor. In order to verify the design and analysis methodology, a canned type 30kW induction motor and an inverter were developed and tested. The motor was designed to have two poles with squirrel cage solid rotor and open slot stator. The motor driver was designed as VVVF inverter to operate both at 900(r.p.m) and at 3600(r.p.m). The calculated design values showed a good agreement with the experimental results. The measured efficiencies of the canned motor and the inverter were 70(%) and 96(%), respectively. A variable reluctance type linear pulse motor (LPM) with double air-gaps for the Control Element Drive Mechanism (CEDM) to lift 100kg was designed, analyzed, manufactured and tested. A converter and a test facility were manufactured to verity the dynamic performance of the LPM. The mover of the LPM was welded with magnetic material(SUS430) and non-magnetic material(SUS304) to get flux path between inner stator and outer stator. The measured thrust force was about 20(%) less than the designed thrust force. As for the rotary stepping motors for CEDM-II, which have transverse flux pattern, three design options were proposed with thrust force density of 8kN/m{sup 2}, 14kN/m{sup 2} and 52kN/m{sup 2} respectively. (author). 31 refs., 219 figs., 60 tabs.

  11. Assessment of the heat carrier movement in the primary coolant circuit by its own momentum

    Energy Technology Data Exchange (ETDEWEB)

    Kadalev, Stoyan, E-mail: kadalev@inrne.bas.bg

    2014-10-15

    Highlights: • We model the heat carrier flow alteration after the circulation pump(s) stop. • The general mathematical model used is described in details. • The model is adapted and applied to a particular example research reactor. • Assessment is presented in detail, step by step with references. • The information provided is enough to apply calculations to another facility. - Abstract: In the presented paper is considered the approach to an assessment of the heat carrier flow alteration in the primary water–water reactor coolant circuit after the circulation pump(s) stop. This topic is highly relevant trough advanced and increased nuclear safety requirements because such a process is observed in case of black-out accident or damaged pump(s). The general mathematical model used is described; enabling preparation of this evaluation adapted and applied to a particular example facility namely a pool type research reactor. The factors influencing to the heat carrier movement by its own momentum are examined. The evaluation measures and includes the factors influencing the heat carrier flow rate from the moment the pump(s) stops down to a negligible value. Assessment is presented in detail, step by step and where needed with references to specific data and/or formulae from reference books to allow repetition of the calculations and/or apply to another facility. The calculations are presented utilizing all necessary data according to the design and technological documentation. No account is given to the pressure of the natural circulation caused by the residual heat generation in the fuel after the reactor scram system extinction of the fission reaction.

  12. Thermal hydraulic characteristics during ingress of coolant and loss of vacuum events in fusion reactors

    Science.gov (United States)

    Takase, K.; Kunugi, T.; Seki, Y.; Akimoto, H.

    2000-03-01

    The thermal hydraulic characteristics in the vacuum vessel (VV) of a fusion reactor under an ingress of coolant event (ICE) and a loss of vacuum event (LOVA) were investigated quantitatively using preliminary experimental apparatuses. In the ICE experiments, pressure rise characteristics in the VV were clarified for experimental parameters of the wall temperature and water temperature and for cases with and without a blowdown tank. In addition, the functional performance of a blowdown tank with and without a water cooling system was examined and it was confirmed that the blowdown tank with a water cooling system is effective for suppressing the pressure rise during the ICE. In the LOVA experiments, the saturation time in the VV from vacuum to atmosphere was investigated for various breach sizes and it was found that the saturation time is in inverse proportion to the breach size. In addition, the characteristics of exchange flow through breaches were clarified for the different breach positions on the VV. It was proven from the experimental results that the exchange flow became a counter-current flow when the breach was positioned on the top of the VV and a stratified flow when it was formed on the side wall of the VV, and that the exchange flow under the stratified flow condition was smoother than that of counter-current flow. On the basis of these results, the severest breach condition in ITER was changed from the top-break case to the side-break case. To predict with high accuracy the thermal hydraulic characteristics during ICEs and LOVAs under ITER conditions, a large scale test facility will be necessary. The current conceptual design of the combined ICE-LOVA test facility with a scaling factor of 1/1000 in comparison with the ITER volume is presented.

  13. Ponderomotive Acceleration in Coronal Loops

    Science.gov (United States)

    Dahlburg, R. B.; Laming, J. M.; Taylor, B. D.; Obenschain, K.

    2016-11-01

    Ponderomotive acceleration has been asserted to be a cause of the first ionization potential (FIP) effect, the well-known enhancement in abundance by a factor of 3-4 over photospheric values of elements in the solar corona with FIP less than about 10 eV. It is shown here by means of numerical simulations that ponderomotive acceleration occurs in solar coronal loops, with the appropriate magnitude and direction, as a “by-product” of coronal heating. The numerical simulations are performed with the HYPERION code, which solves the fully compressible three-dimensional magnetohydrodynamic equations including nonlinear thermal conduction and optically thin radiation. Numerical simulations of coronal loops with an axial magnetic field from 0.005 to 0.02 T and lengths from 25,000 to 75,000 km are presented. In the simulations the footpoints of the axial loop magnetic field are convected by random, large-scale motions. There is a continuous formation and dissipation of field-aligned current sheets, which act to heat the loop. As a consequence of coronal magnetic reconnection, small-scale, high-speed jets form. The familiar vortex quadrupoles form at reconnection sites. Between the magnetic footpoints and the corona the reconnection flow merges with the boundary flow. It is in this region that the ponderomotive acceleration occurs. Mirroring the character of the coronal reconnection, the ponderomotive acceleration is also found to be intermittent.

  14. Dirac Induction for loop groups

    NARCIS (Netherlands)

    Posthuma, H.

    2011-01-01

    Using a coset version of the cubic Dirac operators for affine Lie algebras, we give an algebraic construction of the Dirac induction homomorphism for loop group representations. With this, we prove a homogeneous generalization of the Weyl-Kac character formula and show compatibility with Dirac induc

  15. Five-loop massive tadpoles

    CERN Document Server

    Luthe, T

    2016-01-01

    We provide an update on a long-term project that aims at evaluating massive vacuum integrals at the five-loop frontier, with high precision and in various space-time dimensions. A number of applications are sketched, mainly concerning the determination of anomalous dimensions, for quantum field theories in four, three and two dimensions.

  16. Loop quantum gravity and observations

    CERN Document Server

    Barrau, A

    2014-01-01

    Quantum gravity has long been thought to be completely decoupled from experiments or observations. Although it is true that smoking guns are still missing, there are now serious hopes that quantum gravity phenomena might be tested. We review here some possible ways to observe loop quantum gravity effects either in the framework of cosmology or in astroparticle physics.

  17. Test Report on ISR Double-Loop, Spray-Cooled Inverter

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, John S [ORNL; Coomer, Chester [ORNL; Campbell, Steven L [ORNL; Wiles, Randy H [ORNL; Lowe, Kirk T [ORNL; McFee, Marshall T [ORNL

    2007-02-01

    The Isothermal Systems Research, Inc. (ISR) double-loop, two-phase spray cooling system was designed to use 85 C transmission oil to cool a heat exchanger via a second cooling loop. The heat exchanger condenses the working fluid vapor back to liquid inside a sealed enclosure to allow for continuous spray cooling of electronics. In the ORNL tests, 85 C water/ethylene/glycol (WEG), which has better thermal properties than transmission oil, was substituted for the transmission oil. Because the ISR spray-cooling system requires a second cooling loop, the final inverter might be inherently larger than inverters that do not require a second-loop cooling system. The ISR test setup did not include a dc bus capacitor. Because the insulated gate bipolar transistor (IGBT) conduction test indicated that the ISR test setup could not be properly loaded thermally, no switching tests were conducted. Therefore it was not necessary to attach external capacitors outside the test setup. During load situations not exceeding 400A, the WEG inlet temperature was higher than the WEG outlet temperature. This meant that the 85 C WEG heat exchanger was not cooling the inverter and became a thermal load to the inverter. Only when the load was higher than 400A with a higher coolant temperature and the release valve actuated did the WEG heat exchanger start to cool the 2-phase coolant. The inverter relied strongly on the cooling of the huge aluminum enclosure located inside the ventilation chamber. In a hybrid vehicle, the inverter is situated under the hood, where the dependency on cooling provided by the enclosure may become a problem. The IGBT power dissipation with both sides being spray cooled was around 34 W/cm{sup 2} at 403A, with 995W total IGBT loss at 113.5 C projected junction temperature before the release valve was actuated. The current loading could rise higher than 403 A before reaching the 125 C junction temperature limit if the pressure buildup inside the enclosure could be

  18. High Tc superconducting small loop antenna

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Z.; Mehler, M.J.; Maclean, T.S.M.; Lancaster, M.J.; Gough, C.E. (Univ. of Birmingham (UK)); Alford, N. (I.C.I. Advanced Materials Div., Runcorn (UK))

    1989-12-01

    The improvement in the radiation efficiency of an electrically small loop antenna is analysed when it is fabricated from a superconductor, and experimental results for a liquid nitrogen cooled, ceramic superconducting loop at 450MHz are presented. (orig.).

  19. Crystal packing effects on protein loops.

    Science.gov (United States)

    Rapp, Chaya S; Pollack, Rena M

    2005-07-01

    The effects of crystal packing on protein loop structures are examined by (1) a comparison of loops in proteins that have been crystallized in alternate packing arrangements, and (2) theoretical prediction of loops both with and without the inclusion of the crystal environment. Results show that in a minority of cases, loop geometries are dependent on crystal packing effects. Explicit representation of the crystal environment in a loop prediction algorithm can be used to model these effects and to reconstruct the structures, and relative energies, of a loop in alternative packing environments. By comparing prediction results with and without the inclusion of the crystal environment, the loop prediction algorithm can further be used to identify cases in which a crystal structure does not represent the most stable state of a loop in solution. We anticipate that this capability has implications for structural biology.

  20. Modified Continuous Loop Technique for microvascular anastomosis

    Directory of Open Access Journals (Sweden)

    Kumar Pramod

    2001-01-01

    Full Text Available A modified method of continuous loop technique for microvascular anastomosis is described. The handling of loop is easier & even last suture is placed under vision. This makes the microvascular anastomosis easier and simpler.

  1. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  2. International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

    Directory of Open Access Journals (Sweden)

    N. Aksan

    2008-01-01

    Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

  3. An advanced method for determination of loss of coolant accident in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoodi, R. [Department of Engineering, Shahid Beheshti University, GC, Evin, Tehran (Iran, Islamic Republic of); Shahriari, M., E-mail: m-shahriari@sbu.ac.ir [Department of Engineering, Shahid Beheshti University, GC, Evin, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A. [Department of Engineering, Shahid Beheshti University, GC, Evin, Tehran (Iran, Islamic Republic of)

    2011-06-15

    Highlights: > The considerations of vibration signals are introduced as a new method for determination of accidents directly by detecting of vibration signals without including signals from other components and this is the superiority of the proposed method. > FFT provides an alternate way of representing data. Instead of representing vibration signal amplitude as a function of time, the signal is represented by the amount of information which is contained at different frequencies. > The most of frequencies of structure and fluid coupled are presented in the FFT of structural response and through it the dominant frequency of excitation is obtained. > The Power Spectral Density, a measurement of energy at various frequencies is worked out. MATLAB software is used to convert signals from the time to frequency domain and to obtain PSD of signals. - Abstract: A major objective in reactor design is to provide the capability to withstand a wide range of postulated events without exceeding specified safety limits. Assessment of the consequence of hypothetical loss of coolant accident (LOCA) in primary circuit is an essential element to address fulfilment of acceptance criteria. In addition, finding the position of rupture, one could manage accident in a right direction. In this work, the transient vibration signal from a pipe rupture is used to determine the position of LOCA. A finite element formulation (Galerkin Method) is implemented to include the effect of fluid-structure interaction (FSI). The coupled equations of fluid motion and pipe displacement are solved. The obtained results are in good agreement with published data. Fast Fourier transform (FFT) provides an alternate way of representing data. Instead of representing vibration signal amplitude as a function of time, the signal is represented by the amount of information, which is contained at different frequencies. The most of frequencies of structure and fluid coupled are presented in the FFT of structural

  4. Resumming the POPE at One Loop

    CERN Document Server

    Lam, Ho Tat

    2016-01-01

    The Pentagon Operator Product Expansion represents polygonal Wilson loops in planar $\\mathcal{N}=4$ super Yang-Mills in terms of a series of flux tube excitations for finite coupling. We demonstrate how to re-sum this series at the one loop level for the hexagonal Wilson loop dual to the six-point MHV amplitude. By summing over a series of effective excitations we find expressions which integrate to logarithms and polylogarithms, reproducing the known one-loop result.

  5. Loop Equations in Abelian Gauge Theories

    CERN Document Server

    Di Bartolo, C; Pe~na, F; Bartolo, Cayetano Di; Leal, Lorenzo; Peña, Francisco

    2005-01-01

    The equations obeyed by the vacuum expectation value of the Wilson loop of Abelian gauge theories are considered from the point of view of the loop-space. An approximative scheme for studying these loop-equations for lattice Maxwell theory is presented. The approximation leads to a partial difference equation in the area and length variables of the loop, and certain physically motivated ansatz is seen to reproduce the mean field results from a geometrical perspective.

  6. Estimation of complex permittivity using loop antenna

    DEFF Research Database (Denmark)

    Lenler-Eriksen, Hans-Rudolph; Meincke, Peter

    2004-01-01

    A method for estimating the complex permittivity of materials in the vicinity of a loop antenna is proposed. The method is based on comparing measured and numerically calculated input admittances for the loop antenna.......A method for estimating the complex permittivity of materials in the vicinity of a loop antenna is proposed. The method is based on comparing measured and numerically calculated input admittances for the loop antenna....

  7. Estimation of complex permittivity using loop antenna

    DEFF Research Database (Denmark)

    Lenler-Eriksen, Hans-Rudolph; Meincke, Peter

    2004-01-01

    A method for estimating the complex permittivity of materials in the vicinity of a loop antenna is proposed. The method is based on comparing measured and numerically calculated input admittances for the loop antenna.......A method for estimating the complex permittivity of materials in the vicinity of a loop antenna is proposed. The method is based on comparing measured and numerically calculated input admittances for the loop antenna....

  8. Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm{sup 2}, simulated with RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: gdgian@ipen.br, E-mail: borges.em@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

  9. The aerodynamic effects of wheelspace coolant injection into the mainstream flow of a high pressure gas turbine

    Science.gov (United States)

    McLean, Christopher Elliot

    Modern gas turbine engines operate with mainstream gas temperatures exceeding 1450°C in the high-pressure turbine stage. Unlike turbine blades, rotor disks and other internal components are not designed to withstand the extreme temperatures found in mainstream flow. In modern gas turbines, cooling air is pumped into the wheelspace cavities to prevent mainstream gas ingestion and then exits through a seal between the rotor and the nozzle guide vane (NGV) thereby mixing with the mainstream flow. The primary purpose for the wheelspace cooling air is the cooling of the turbine wheelspace. However, secondary effects arise from the mixing of the spent cooling air with the mainstream flow. The exiting cooling air is mixed with the hot mainstream flow effecting the aerodynamic and performance characteristics of the turbine stage. The physics underlying this mixing process and its effects on stage performance are not yet fully understood. The relative aerodynamic and performance effects associated with rotor - NGV gap coolant injections were investigated in the Axial Flow Turbine Research Facility (AFTRF) of the Center for Gas Turbines and Power of The Pennsylvania State University. This study quantifies the secondary effects of the coolant injection on the aerodynamic and performance character of the turbines main stream flow for root injection, radial cooling, and impingement cooling. Measurement and analysis of the cooling effects were performed in both stationary and rotational frames of reference. The AFTRF is unique in its ability to perform long duration cooling measurements in the stationary and rotating frames. The effects of wheelspace coolant mixing with the mainstream flow on total-to-total efficiency, energy transport, three dimensional velocity field, and loading coefficient were investigated. Overall, it was found that a small quantity (1%) of cooling air can have significant effects on the performance character and exit conditions of the high pressure stage

  10. Effect of external turbulence on the efficiency of film cooling with coolant injection into a transverse trench

    Science.gov (United States)

    Khalatov, A. A.; Panchenko, N. A.; Severin, S. D.

    2017-09-01

    Film cooling is among the basic methods used for thermal protection of blades in modern high-temperature gas turbines. Results of computer simulation of film cooling with coolant injection via a row of conventional inclined holes or a row of holes in a trench are presented in this paper. The ANSYS CFX 14 commercial software package was used for CFD-modeling. The effect is studied of the mainstream turbulence on the film cooling efficiency for the blowing ratio range between 0.6 and 2.3 and three different turbulence intensities of 1, 5, and 10%. The mainstream velocity was 150 and 400 m/s, while the temperatures of the mainstream and the injected coolant were 1100 and 500°C, respectively. It is demonstrated that, for the coolant injection via one row of trenched holes, an increase in the mainstream turbulence intensity reduces the film cooling efficiency in the entire investigated range of blowing ratios. It was revealed that freestream turbulence had varied effects on the film cooling efficiency depending on the blowing ratio and mainstream velocity in a blade channel. Thus, an increase in the mainstream turbulence intensity from 1 to 10% decreases the surface-averaged film cooling efficiency by 3-10% at a high mainstream velocity (400 m/s) in the blade channel and by 12-23% at a moderate velocity (of 150 m/s). Here, lower film cooling efficiencies correspond to higher blowing ratios. The effect of mainstream turbulence intensity on the film cooling efficiency decreases with increasing the mainstream velocity in the modeled channel for both investigated configurations.

  11. Characterization of Industrial Coolant Fluids and Continuous Ageing Monitoring by Wireless Node-Enabled Fiber Optic Sensors

    DEFF Research Database (Denmark)

    Sachat, Alexandros El; Meristoudi, Anastasia; Markos, Christos

    2017-01-01

    Environmentally robust chemical sensors for monitoring industrial processes or infrastructures are lately becoming important devices in industry. Low complexity and wireless enabled characteristics can offer the required flexibility for sensor deployment in adaptable sensing networks for continuous...... of sensors allowed their integration in an autonomous wireless sensing node, thus enabling the future use of the demonstrated platform in wireless sensor networks for a variety of industrial and environmental monitoring applications.......-gel derived pH sensitive coatings. The performances of the developed sensors were characterized and compared, towards their coolants' ageing monitoring capability, proving their efficiency in such a demanding application scenario and harsh industrial environment. The operating characteristics of this type...

  12. Chemical and radiolytical characterization of perfluorocarbon fluids used as coolants for LHC experiments : radiolysis effects in perfluorohexane fluids.

    CERN Document Server

    Ilie, Soran; Teissandier, B; CERN. Geneva. TS Department

    2007-01-01

    Perfluorohexane fluids, used as coolants within High Energy Physics Detectors in the Large Hadrons Collider (LHC) at CERN, were irradiated using gammas 60Co and characterized using different analytical techniques. The aim of this work was the assessment of radiation induced effects as a function of the chemical nature of these fluids and their impurity content. Were evidenced the radioinduced polymers and acidity, as well as different chemical by-products. Purification tests and measurements were carried out on different irradiated fluid samples to assess the efficiency of such purification treatments in view of their re-use in the HEP detector cooling systems.

  13. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  14. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  15. Effect of Grinding Process Parameters on Surface Area Roughness of Glass fibre Reinforced Composite Laminate under Dry and Coolant Environment

    Directory of Open Access Journals (Sweden)

    P. Chockalingam

    2016-04-01

    Full Text Available This paper presents a comparative study on dry and wet grinding of chopped strand mat glass fibre reinforced polymer laminates using an alumina wheel. Investigations were performed to study the impact of the grinding parameters, namely feed, speed, and depth of cut on grinding force ratio and surface area roughness. Effective grinding parameters were sought in this study to maximize grinding force ratio and minimize surface area roughness. Test results show that coolant helped to decrease surface area roughness, but inevitably reduced the grinding force ratio in some cases. These findings lead to economic machining solution for optimum grinding conditions in grinding composite laminates.

  16. Polyhedra in loop quantum gravity

    CERN Document Server

    Bianchi, Eugenio; Speziale, Simone

    2010-01-01

    Interwiners are the building blocks of spin-network states. The space of intertwiners is the quantization of a classical symplectic manifold introduced by Kapovich and Millson. Here we show that a theorem by Minkowski allows us to interpret generic configurations in this space as bounded convex polyhedra in Euclidean space: a polyhedron is uniquely described by the areas and normals to its faces. We provide a reconstruction of the geometry of the polyhedron: we give formulas for the edge lengths, the volume and the adjacency of its faces. At the quantum level, this correspondence allows us to identify an intertwiner with the state of a quantum polyhedron, thus generalizing the notion of quantum tetrahedron familiar in the loop quantum gravity literature. Moreover, coherent intertwiners result to be peaked on the classical geometry of a polyhedron. We discuss the relevance of this result for loop quantum gravity. In particular, coherent spin-network states with nodes of arbitrary valence represent a collection...

  17. Nucleosome repositioning via loop formation

    CERN Document Server

    Kulic, M L

    2002-01-01

    Active (catalysed) and passive (intrinsic) nucleosome repositioning is known to be a crucial event during the transcriptional activation of certain eucaryotic genes. Here we consider theoretically the intrinsic mechanism and study in detail the energetics and dynamics of DNA-loop-mediated nucleosome repositioning, as previously proposed by Schiessel et al. (H. Schiessel, J. Widom, R. F. Bruinsma, and W. M. Gelbart. 2001. {\\it Phys. Rev. Lett.} 86:4414-4417). The surprising outcome of the present study is the inherent nonlocality of nucleosome motion within this model -- being a direct physical consequence of the loop mechanism. On long enough DNA templates the longer jumps dominate over the previously predicted local motion, a fact that contrasts simple diffusive mechanisms considered before. The possible experimental outcome resulting from the considered mechanism is predicted, discussed and compared to existing experimental findings.

  18. The Statistical Loop Analyzer (SLA)

    Science.gov (United States)

    Lindsey, W. C.

    1985-01-01

    The statistical loop analyzer (SLA) is designed to automatically measure the acquisition, tracking and frequency stability performance characteristics of symbol synchronizers, code synchronizers, carrier tracking loops, and coherent transponders. Automated phase lock and system level tests can also be made using the SLA. Standard baseband, carrier and spread spectrum modulation techniques can be accomodated. Through the SLA's phase error jitter and cycle slip measurements the acquisition and tracking thresholds of the unit under test are determined; any false phase and frequency lock events are statistically analyzed and reported in the SLA output in probabilistic terms. Automated signal drop out tests can be performed in order to trouble shoot algorithms and evaluate the reacquisition statistics of the unit under test. Cycle slip rates and cycle slip probabilities can be measured using the SLA. These measurements, combined with bit error probability measurements, are all that are needed to fully characterize the acquisition and tracking performance of a digital communication system.

  19. Loop Diuretics in Clinical Practice.

    Science.gov (United States)

    Oh, Se Won; Han, Sang Youb

    2015-06-01

    Diuretics are commonly used to control edema across various clinical fields. Diuretics inhibit sodium reabsorption in specific renal tubules, resulting in increased urinary sodium and water excretion. Loop diuretics are the most potent diuretics. In this article, we review five important aspects of loop diuretics, in particular furosemide, which must be considered when prescribing this medicine: (1) oral versus intravenous treatment, (2) dosage, (3) continuous versus bolus infusion, (4) application in chronic kidney disease patients, and (5) side effects. The bioavailability of furosemide differs between oral and intravenous therapy. Additionally, the threshold and ceiling doses of furosemide differ according to the particular clinical condition of the patient, for example in patients with severe edema or chronic kidney disease. To maximize the efficiency of furosemide, a clear understanding of how the mode of delivery will impact bioavailability and the required dosage is necessary.

  20. Deconfinement and virtual quark loops

    Science.gov (United States)

    Çelik, T.; Engels, J.; Satz, H.

    1983-12-01

    We calculate paer Monte Carlo evaluation on an 83 × 3 lattice the energy density ɛG of the gluon sector of QCD, including virtual quark loops up to the fourth power in the hopping parameter expansion. For light quarks of one flavour, we observe at T/ΛL 95 +/- 10 a rapid variation of ɛG in T, accompanied by strong fluctuations from iteration to iteration. as clear signal of the deconfinement transition.

  1. Quantum Reduced Loop Gravity and the foundation of Loop Quantum Cosmology

    CERN Document Server

    Alesci, Emanuele

    2016-01-01

    Quantum Reduced Loop Gravity is a promising framework for linking Loop Quantum Gravity and the effective semiclassical dynamics of Loop Quantum Cosmology. We review its basic achievements and its main perspectives, outlining how it provides a quantum description of the Universe in terms of a cuboidal graph which constitutes the proper framework for applying loop techniques in a cosmological setting.

  2. Ponderomotive Acceleration in Coronal Loops

    CERN Document Server

    Dahlburg, R B; Taylor, B D; Obenschain, K

    2016-01-01

    Ponderomotive acceleration has been asserted to be a cause of the First Ionization Potential (FIP) effect, the by now well known enhancement in abundance by a factor of 3-4 over photospheric values of elements in the solar corona with FIP less than about 10 eV. It is shown here by means of numerical simulations that ponderomotive acceleration occurs in solar coronal loops, with the appropriate magnitude and direction, as a "byproduct" of coronal heating. The numerical simulations are performed with the HYPERION code, which solves the fully compressible three-dimensional magnetohydrodynamic equations including nonlinear thermal conduction and optically thin radiation. Numerical simulations of a coronal loops with an axial magnetic field from 0.005 Teslas to 0.02 Teslas and lengths from 25000 km to 75000 km are presented. In the simulations the footpoints of the axial loop magnetic field are convected by random, large-scale motions. There is a continuous formation and dissipation of field-aligned current sheets...

  3. Modeling Phase-Locked Loops Using Verilog

    Science.gov (United States)

    2007-11-01

    a charge pump, the phase detector has a tri-state output that can drive a opamp loop filter directly. This signal is conditioned by the charge pump...then it can directly drive an opamp based loop filter. Most loop filters are based upon an integrator loop. The integrator loop filter is advantageous...replaced with an accumulator. The opamp circuit can be replaced by a digital filter using Z-transform theory z=exp(jwT), where T is the sampling

  4. Dissolved oxygen control and monitoring implementation in the liquid lead bismuth eutectic loop: HELIOS

    Science.gov (United States)

    Nam, Hyo On; Lim, Jun; Han, Dong Yoon; Hwang, Il Soon

    2008-06-01

    A 12 m tall LBE coolant loop, named as HELIOS, has been developed by thermal-hydraulic scaling of the PEACER-300MWe. Thermo-hydraulic experiment and materials test are the principal purposes of HELIOS operation. In this study, an yttria stabilized zirconia (YSZ) based oxygen sensor that was hermetically sealed for long-term applications using the electromagnetically swaged metal-ceramic joining method, have been developed for high temperature oxygen control application over a long period of time. The rugged electrode design has been calibrated to absolute metal-oxide equilibrium by using a first principle of detecting pure metal-oxide transition using electrochemical impedance spectroscopy (EIS). During the materials tests in HELIOS, dissolved oxygen concentration was administered at the intended condition of 10 -6 wt% by direct gas bubbling with Ar + 4%H 2, Ar + 5%O 2 and/or pure Ar while corrosion tests were conducted for up to 1000 h with inspection after each 333 h. During the total 1000 h corrosion test, oxygen concentration was measured by oxygen sensor. The result confirmed that the direct gas bubbling method is a viable and practical option for controlling oxygen concentration in large loops including HELIOS.

  5. Bellows-Type Accumulators for Liquid Metal Loops of Space Reactor Power Systems

    Science.gov (United States)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2006-01-01

    In many space nuclear power systems, the primary and/or secondary loops use liquid metal working fluids, and require accumulators to accommodate the change in the liquid metal volume and maintain sufficient subcooling to avoid boiling. This paper developed redundant and light-weight bellows-type accumulators with and without a mechanical spring, and compared the operating condition and mass of the accumulators for different types of liquid metal working fluids and operating temperatures: potassium, NaK-78, sodium and lithium loops of a total capacity of 50 liters and nominal operating temperatures of 840 K, 860 K, 950 K and 1340 K, respectively. The effects of using a mechanical spring and different structural materials on the design, operation and mass of the accumulators are also investigated. The structure materials considered include SS-316, Hastelloy-X, C-103 and Mo-14Re. The accumulator without a mechanical spring weighs 23 kg and 40 kg for a coolant subcooling of 50 K and 100 K, respectively, following a loss of the fill gas. The addition of a mechanical spring comes with a mass penalty, in favor of higher redundancy and maintaining a higher liquid metal subcooling.

  6. The space of states of quantum gravity in terms of loops and extended loops some remarks

    CERN Document Server

    Di Bartolo, C; Griego, J R; Pullin, J; Di Bartolo, Cayetano; Gambini, Rodolfo; Griego, Jorge; Pullin, Jorge

    1995-01-01

    This article reviews the status of several solutions to all the constraints of quantum gravity that have been proposed in terms of loops and extended loops. We discuss pitfalls of several of the results and in particular discuss the issues of covariance and regularization of the constraints in terms of extended loops. We also propose a formalism for ``thickened out loops'' which does not face the covariance problems of extended loops and may allow to regularize expressions in a consistent manner.

  7. Hyperstaticity and loops in frictional granular packings

    Science.gov (United States)

    Tordesillas, Antoinette; Lam, Edward; Metzger, Philip T.

    2009-06-01

    The hyperstatic nature of granular packings of perfectly rigid disks is analyzed algebraically and through numerical simulation. The elementary loops of grains emerge as a fundamental element in addressing hyperstaticity. Loops consisting of an odd number of grains behave differently than those with an even number. For odd loops, the latent stresses are exterior and are characterized by the sum of frictional forces around each loop. For even loops, the latent stresses are interior and are characterized by the alternating sum of frictional forces around each loop. The statistics of these two types of loop sums are found to be Gibbsian with a "temperature" that is linear with the friction coefficient μ when μ<1.

  8. The extended loop representation of quantum gravity

    CERN Document Server

    Di Bartolo, C; Griego, J R

    1995-01-01

    A new representation of Quantum Gravity is developed. This formulation is based on an extension of the group of loops. The enlarged group, that we call the Extended Loop Group, behaves locally as an infinite dimensional Lie group. Quantum Gravity can be realized on the state space of extended loop dependent wavefunctions. The extended representation generalizes the loop representation and contains this representation as a particular case. The resulting diffeomorphism and hamiltonian constraints take a very simple form and allow to apply functional methods and simplify the loop calculus. In particular we show that the constraints are linear in the momenta. The nondegenerate solutions known in the loop representation are also solutions of the constraints in the new representation. The practical calculation advantages allows to find a new solution to the Wheeler-DeWitt equation. Moreover, the extended representation puts in a precise framework some of the regularization problems of the loop representation. We sh...

  9. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

    Directory of Open Access Journals (Sweden)

    V. Sánchez

    2010-01-01

    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  10. Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

    Directory of Open Access Journals (Sweden)

    Lefèvre Grégory

    2012-01-01

    Full Text Available In the primary circuit of pressurized water reactors (PWR, the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek theory and used as such to interpret this industrial phenomenon.

  11. Cold neutron tomography of annular coolant flow in a double subchannel model of a boiling water reactor

    Science.gov (United States)

    Kickhofel, J. L.; Zboray, R.; Damsohn, M.; Kaestner, A.; Lehmann, E. H.; Prasser, H.-M.

    2011-09-01

    Dryout of the liquid coolant film on fuel pins at the top of boiling water reactor (BWR) cores constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is a limiting factor in the thermal power, and therefore the economy, of BWRs. Ongoing research on multiphase annular flow, specifically the liquid film thickness, is fundamental not only to nuclear reactor safety and operation but also to that of evaporators, condensers, and pipelines in a general industrial context. We have performed cold neutron tomography of adiabatic air water annular flow in a scaled up model of the subchannel geometry found in BWR fuel assemblies today. All imaging has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institut in Switzerland. Neutron tomography is shown to excel in investigating the interactions of air water two phase flows with spacer vanes of different geometry. The high resolution, high contrast measurements provide spatial distributions of the coolant on top of the surfaces of the spacer, including the vanes, and in the subchannel downstream of the spacers.

  12. Analytical study on creep behavior of a tube of coolant piping system in nuclear power plant. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, Noriyuki [Kyushu Univ., Fukuoka (Japan); Hagihara, Seiya [Saga Univ., Saga (Japan); Chino, Eiichi; Maeda, Akio [MRI Systems Inc., Tokyo (Japan); Maruyama, Yu; Hashimoto, Kazuichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-10-01

    During severe accident of a light water reactor (LWR), reactor coolant piping would be damaged when the piping is subjected to high internal pressure and very high temperature due to heat transfer from high-temperature gas and decay heat from wall-deposited fission product (FP), both from degraded core. In such a case, high-temperature fast creep deformation could be the main cause for the pipe failure. For the evaluation of piping integrity during severe accidents, a method to predict such high-temperature fast creep deformation should be developed, using a creep constitutive equation considering tertiary creep behavior which has not been considered well in the pipe failure analyses. In this study, a creep constitutive equation was developed first based on the Kachanov-Ravotnov isotropic damage rule that considers the tertiary creep behavior. JAERI creep tensile test data for both nuclear-grade cold-drawn SUS316N and hot-extruded SUS316 materials were used to determine coefficients of the developed constitutive equation. Using the developed constitutive equation, finite element analyses were performed for local creep deformation of coolant piping under two temperature conditions: uniform temperature and temperature gradient. The analytical results indicated the damage variable being integrated following the evolution of creep damage can indicate pipe wall internal damage condition quantitatively. The damage variable was confirmed further to be able to reproduce the observation in JAERI piping failure tests, that is, pipe failure from the wall outside. (author)

  13. The development of robotic system for inspecting and repairing NPP primary coolant system of high-level radioactive environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Kim, Ki Ho; Jung, Seung Ho; Kim, Byung Soo; Hwang, Suk Yeoung; Kim, Chang Hoi; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Cho, Jai Wan; Lee, Jae Kyung; Lee, Yong Deok

    1997-07-01

    This project aims at developing a robotic system to automatically handle inspection and maintenance of NPP safety-related facilities in high-level radioactive environment. This robotic system under development comprises two robots depending on application fields - a mobile robot and multi-functional robot. The mobile robot is designed to be used in the area of primary coolant system during the operation of NPP. This robot enables to overcome obstacles and perform specified tasks in unstructured environment. The multi-functional robot is designed for performing inspection and maintenance tasks of steam generator and nuclear reactor vessel during the overhaul periods of NPP. Nuclear facilities can be inspected and repaired all the time by use of both the mobile robot and the multi-functional robot. Human operator, by teleoperation, monitors the movements of such robots located at remote task environment via video cameras and controls those remotely generating desired commands via master manipulator. We summarize the technology relating to the application of the mobile robot to primary coolant system environment, the applicability of the mobile robot through 3D graphic simulation, the design of the mobile robot, the design of its radiation-hardened controller. We also describe the mechanical design, modeling, and control system of the multi-functional robot. Finally, we present the design of the force-reflecting master and the modeling of virtual task environment for a training simulator. (author). 47 refs., 16 tabs., 43 figs.

  14. Characterization of Industrial Coolant Fluids and Continuous Ageing Monitoring by Wireless Node—Enabled Fiber Optic Sensors

    Science.gov (United States)

    El Sachat, Alexandros; Meristoudi, Anastasia; Markos, Christos; Sakellariou, Andreas; Papadopoulos, Aggelos; Katsikas, Serafim; Riziotis, Christos

    2017-01-01

    Environmentally robust chemical sensors for monitoring industrial processes or infrastructures are lately becoming important devices in industry. Low complexity and wireless enabled characteristics can offer the required flexibility for sensor deployment in adaptable sensing networks for continuous monitoring and management of industrial assets. Here are presented the design, development and operation of a class of low cost photonic sensors for monitoring the ageing process and the operational characteristics of coolant fluids used in an industrial heavy machinery infrastructure. The chemical, physical and spectroscopic characteristics of specific industrial-grade coolant fluids were analyzed along their entire life cycle range, and proper parameters for their efficient monitoring were identified. Based on multimode polymer or silica optical fibers, wide range (3–11) pH sensors were developed by employing sol-gel derived pH sensitive coatings. The performances of the developed sensors were characterized and compared, towards their coolants’ ageing monitoring capability, proving their efficiency in such a demanding application scenario and harsh industrial environment. The operating characteristics of this type of sensors allowed their integration in an autonomous wireless sensing node, thus enabling the future use of the demonstrated platform in wireless sensor networks for a variety of industrial and environmental monitoring applications. PMID:28287488

  15. Radiogenic Lead with Dominant Content of 208Pb: New Coolant and Neutron Moderator for Innovative Nuclear Facilities

    Directory of Open Access Journals (Sweden)

    A. N. Shmelev

    2011-01-01

    Full Text Available As a rule materials of small atomic weight (light and heavy water, graphite, and so on are used as neutron moderators and reflectors. A new very heavy atomic weight moderator is proposed—radiogenic lead consisting mainly of isotope 208Pb. It is characterized by extremely low neutron radiative capture cross-section (0.23 mbarn for thermal neutrons, i.e., less than that for graphite and deuterium and highest albedo of thermal neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in a fast reactor. This can increase safety of the fast reactors and reduce as well requirements pertaining to the fuel fabrication technology. Radiogenic lead with high 208Pb content as a liquid-metal coolant of fast reactors helps to achieve a favorable (negative reactivity coefficient on coolant temperature. It is noteworthy that radiogenic lead with high 208Pb content may be extracted from thorium (as well as thorium-uranium ores without isotope separation. This has been confirmed experimentally by the investigations performed at San Paulo University, Brazil.

  16. Recommendations on selecting the closing relations for calculating friction pressure drop in the loops of nuclear power stations equipped with VVER reactors

    Science.gov (United States)

    Alipchenkov, V. M.; Belikov, V. V.; Davydov, A. V.; Emel'yanov, D. A.; Mosunova, N. A.

    2013-05-01

    Closing relations describing friction pressure drop during the motion of two-phase flows that are widely applied in thermal-hydraulic codes and in calculations of the parameters characterizing the flow of water coolant in the loops of reactor installations used at nuclear power stations and in other thermal power systems are reviewed. A new formula developed by the authors of this paper is proposed. The above-mentioned relations are implemented in the HYDRA-IBRAE thermal-hydraulic computation code developed at the Nuclear Safety Institute of the Russian Academy of Sciences. A series of verification calculations is carried out for a wide range of pressures, flowrates, and heat fluxes typical for transient and emergency operating conditions of nuclear power stations equipped with VVER reactors. Advantages and shortcomings of different closing relations are revealed, and recommendations for using them in carrying out thermal-hydraulic calculations of coolant flow in the loops of VVER-based nuclear power stations are given.

  17. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  18. Computational analysis of coolant mixing in subassembly and hot pool of an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Velusamy, K.; Kasinathan, N.; Clement Ravichandar, S.; Selvaraj, P.; Ghosh, D.; Chellapandi, P.; Chetal, S.C. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam - 603 102 (India)

    2005-07-01

    Full text of publication follows: The reactor core of a typical Liquid Metal cooled Fast Breeder Reactor (LMFBR) comprises of several fuel, blanket, storage, reflector and shielding subassemblies (SA) arranged in a hexagonal geometry. Each fuel subassembly (FSA) comprises of 217 fuel pins packed with pellets of fissile and fertile materials. The wire wrapped pins are arranged in triangular pitch within a hexagonal wrapper. Due to the non uniform flow resistance across the cross section, sodium velocity is non uniform within SA. This non uniform velocity distribution coupled with radial power profile causes the sodium coolant to come out of the pin bundles of SA with a non uniform temperature profile. From the fuel pin bundle top, sodium flows axially through various regions comprising of a sodium plenum, shielding bundle region and SA head and adapter assembly. These regions contribute in the mixing of sodium in SA. Due to the non uniform temperature distribution at FSA top, the temperature profile at the possible locations of core monitoring thermocouple (TC) would also be non uniform. During a power campaign of core, FSA top drift outwards with respect to TC location due to irradiation bowing. This causes the temperature recorded by the TC to change in accordance with the temperature of sodium stream washing it from time to time during the power campaign. Apart from this, the non uniform temperature at the FSA exits can also lead to fluctuation in the temperature of sodium at TC locations. Knowledge of these phenomena in hot pool is essential in fixing the SCRAM threshold on the parameters that are derived from the measurements made by these TC. Analyses to study the above phenomena have been carried out in three stages. In the first stage, velocity and temperature distributions in the wire wrap bundle region of FSA have been obtained through subchannel analysis using the SUPER ENERGY code. In the second stage, 3D mixing of sodium in FSA in the region above the

  19. Corrosion test using commercial FMS and developed ARAA in the Experimental loop for liquid breeder

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, J. S.; Kim, S. K.; Lee, E. H.; Jin, H. G.; Lee, D. W.; Jung, Y. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, H. G. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A liquid-type TBM, a Helium Cooled Molten Lithium (HCML) TBM, is one of candidate TBM for KO Demos. In this concept, helium (He) and liquid lithium (Li) were used as a coolant and a breeder, respectively. However, according to our strategy for developing the liquid breeder TBM and its more relevant DEMO concept, liquid breeders not only considered liquid lithium but also lead-lithium (PbLi). An Experimental Loop for Liquid breeder (ELLI) was constructed for the purpose of validating the electromagnetic (EM) pump design, which designed and fabricated by ourselves; testing the effects of magneto-hydro-dynamics (MHD); and investigating the compatibility of PbLi using structural materials such as ferritic martensitic steel. The picture of the experimental loop is shown in Fig. 1. The performance test on each component such as heaters, the control systems for heating the loop were performed and the characteristic tests with a magnet and the EM pump were carried out. A corrosion tests using ELLI were performed with grade 91 FMS steel and ARAA during 250 hours. The corrosion tests specimens using developed commercial-scale ARAA-1 will be carried out in the ELLI loop using EM pump to investigate the corrosion behavior of ARAA-1, and the test results will be compared the previous corrosion tests of FMS corrosion specimens. A commercial FMS with grade 91 and developed ARAA steel was used for corrosion test-specimens to compare the corrosion characteristics in the flowing Pbli loop. Long term operation performance of the ELLI loop was conformed during 250 h corrosion tests. During the two separate experiments, an EM pump was operated for 250 h with a speed of 0.16 m/s (0.95 Kg/s) for corrosion tests. Mass of samples was decreased about 0.006% for ARAA-sample and 0.009% for Gr.91-sample, which corresponded to 0.22 g/m2 and 0.34 g/m2 of a corrosion attack, respectively. Unfortunately, it was unavailable to compare these weight losses with previous reports because of the

  20. Investigating hydrodynamic characteristics and peculiarities of the coolant flow behind a spacer grid of a fuel rod assembly of the floating nuclear power unit

    Science.gov (United States)

    Dmitriev, S. M.; Doronkov, D. V.; Legchanov, M. A.; Pronin, A. N.; Solncev, D. N.; Sorokin, V. D.; Hrobostov, A. E.

    2016-05-01

    The results of experimental investigations of local hydrodynamics of a coolant flow in fuel rod assembly (FA) of KLT-40C reactor behind a plate spacer grid have been presented. The investigations were carried out on an aerodynamic rig using the gas-phase diffusive tracer test. An analysis of spatial distribution of absolute flow velocity projections and distribution of tracer concentration allowed specifying a coolant flow pattern behind the plate spacer grid of the FA. On the basis of obtained experimental data the recommendations were provided to specify procedures for determining the coolant flow rates for the programs of cell-wise calculation of a core zone of KLT-40C reactor. Investigation results were accepted for the practical use in JSC "OKBM Afrikantov" to assess heat engineering reliability of cores of KLT-40C reactor and were included in a database for verification of CFD programs (CFD-codes).

  1. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  2. Working regime identification for natural circulation loops by comparative thermalhydraulic analyses with three fluids under identical operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Milan K.S.; Basu, Dipankar N., E-mail: dipankar.n.basu@gmail.com

    2015-11-15

    Highlights: • Thermalhydraulic analyses of NCL to justify the use of supercritical condition. • Mass flow rate of supercritical loop increases with heater power till a maxima. • Supercritical loop suffer from HTD beyond the maxima with jump in fluid temperature. • HTD is pronounced at higher sink temperatures and pressures just above critical. • Supercritical CO{sub 2} is preferred fluid till the HTD and single-phase water afterwards. - Abstract: Computational investigation for comparative thermalhydraulic analyses of rectangular natural circulation loops is performed to propose a guideline for selecting the working fluid and nature of the loop, subcritical or supercritical, under identical levels of operating parameters like pressure, heating power and coolant temperature. A 3-d uniform-diameter loop geometry is developed with horizontal heating and cooling. Heating is provided in constant heat flux mode, whereas cooling is through a constant temperature sink. Due to favourable thermophysical properties and environmental conformity, water, CO{sub 2} and R134a are selected as possible working fluids. Operational parameters are set so as to have sub- to supercritical condition for CO{sub 2}, supercritical for R134a and single-phase liquid for water. Mass flow rate for supercritical fluid rapidly increases with heater power, when the fluid is allowed to cross the pseudocritical point during its passage through the heater, and exhibits a maxima. Drastic fall in mass flow rate can be observed beyond the maxima, accompanied by a jump in maximum fluid temperature and a rapid decline in sink-side heat transfer coefficient. That can be identified as heat transfer deterioration in supercritical natural circulation loops, a highly undesirable situation from loop safety point of view. Allowable working range of heater power can be enhanced by increasing system pressure and decreasing sink temperature. For any specified set of operating conditions, CO{sub 2}-based

  3. FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G. [and others

    1994-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m{sup 2}-s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ``hot channel``. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m{sup 2}, a mass flux range of 8 to 28 Mg/m{sup 2}-s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ``soft`` a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author`s knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure.

  4. Advanced neutron source reactor thermal-hydraulic test loop facility description

    Energy Technology Data Exchange (ETDEWEB)

    Felde, D.K.; Farquharson, G.; Hardy, J.H.; King, J.F.; McFee, M.T.; Montgomery, B.H.; Pawel, R.E.; Power, B.H.; Shourbaji, A.A.; Siman-Tov, M.; Wood, R.J.; Yoder, G.L.

    1994-02-01

    The Thermal-Hydraulic Test Loop (THTL) is a facility for experiments constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory. The ANSR is both cooled and moderated by heavy water and uses uranium silicide fuel. The core is composed of two coaxial fuel-element annuli, each of different diameter. There are 684 parallel aluminum-clad fuel plates (252 in the inner-lower core and 432 in the outer-upper core) arranged in an involute geometry that effectively creates an array of thin rectangular flow channels. Both the fuel plates and the coolant channels are 1.27 mm thick, with a span of 87 mm (lower core), 70 mm (upper core), and 507-mm heated length. The coolant flows vertically upwards at a mass flux of 27 Mg/m{sup 2}s (inlet velocity of 25 m/s) with an inlet temperature of 45{degrees}C and inlet pressure of 3.2 MPa. The average and peak heat fluxes are approximately 6 and 12 MW/m{sup 2}, respectively. The availability of experimental data for both flow excursion (FE) and true critical heat flux (CHF) at the conditions applicable to the ANSR is very limited. The THTL was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of thermal limits under the expected ANSR thermal-hydraulic conditions. For these experimental studies, the involute-shaped fuel plates of the ANSR core with the narrow 1.27-mm flow gap are represented by a narrow rectangular channel. Tests in the THTL will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat-transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat-transfer coefficients, (4) two-phase pressure-drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) CHF limits.

  5. A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Husain, A.

    2003-02-27

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.

  6. Singularities in loop quantum cosmology.

    Science.gov (United States)

    Cailleteau, Thomas; Cardoso, Antonio; Vandersloot, Kevin; Wands, David

    2008-12-19

    We show that simple scalar field models can give rise to curvature singularities in the effective Friedmann dynamics of loop quantum cosmology (LQC). We find singular solutions for spatially flat Friedmann-Robertson-Walker cosmologies with a canonical scalar field and a negative exponential potential, or with a phantom scalar field and a positive potential. While LQC avoids big bang or big rip type singularities, we find sudden singularities where the Hubble rate is bounded, but the Ricci curvature scalar diverges. We conclude that the effective equations of LQC are not in themselves sufficient to avoid the occurrence of curvature singularities.

  7. Loop equations from differential systems

    CERN Document Server

    Eynard, Bertrand; Marchal, Olivier

    2016-01-01

    To any differential system $d\\Psi=\\Phi\\Psi$ where $\\Psi$ belongs to a Lie group (a fiber of a principal bundle) and $\\Phi$ is a Lie algebra $\\mathfrak g$ valued 1-form on a Riemann surface $\\Sigma$, is associated an infinite sequence of "correlators" $W_n$ that are symmetric $n$-forms on $\\Sigma^n$. The goal of this article is to prove that these correlators always satisfy "loop equations", the same equations satisfied by correlation functions in random matrix models, or the same equations as Virasoro or W-algebra constraints in CFT.

  8. Cygnus Loop: A double bubble?

    CERN Document Server

    West, J; Reichardt, I; Stil, J; Kothes, R; Jaffe, T

    2016-01-01

    The Cygnus Loop is a well-studied supernova remnant (SNR) that has been observed across the electromagnetic spectrum. Although widely believed to be an SNR shell with a blow- out region in the south, we consider the possibility that this object is two SNRs projected along the same line-of-sight by using multi-wavelength images and modelling. Our results show that a model of two objects including some overlap region/interaction between the two objects has the best match to the observed data.

  9. Closed-loop neuromorphic benchmarks

    CSIR Research Space (South Africa)

    Stewart

    2015-11-01

    Full Text Available Benchmarks   Terrence C. Stewart 1* , Travis DeWolf 1 , Ashley Kleinhans 2 , Chris Eliasmith 1   1 University of Waterloo, Canada, 2 Council for Scientific and Industrial Research, South Africa   Submitted to Journal:   Frontiers in Neuroscience   Specialty... the study was exempt from ethical approval procedures.) Did the study presented in the manuscript involve human or animal subjects: No I v i w 1Closed-loop Neuromorphic Benchmarks Terrence C. Stewart 1,∗, Travis DeWolf 1, Ashley Kleinhans 2 and Chris...

  10. Loop Virasoro Lie conformal algebra

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Henan, E-mail: wuhenanby@163.com; Chen, Qiufan; Yue, Xiaoqing [Department of Mathematics, Tongji University, Shanghai 200092 (China)

    2014-01-15

    The Lie conformal algebra of loop Virasoro algebra, denoted by CW, is introduced in this paper. Explicitly, CW is a Lie conformal algebra with C[∂]-basis (L{sub i} | i∈Z) and λ-brackets [L{sub i} {sub λ} L{sub j}] = (−∂−2λ)L{sub i+j}. Then conformal derivations of CW are determined. Finally, rank one conformal modules and Z-graded free intermediate series modules over CW are classified.

  11. Using the Meeting Graph Framework to Minimise Kernel Loop Unrolling for Scheduled Loops

    Science.gov (United States)

    Bachir, Mounira; Gregg, David; Touati, Sid-Ahmed-Ali

    This paper improves our previous research effort [1] by providing an efficient method for kernel loop unrolling minimisation in the case of already scheduled loops, where circular lifetime intervals are known. When loops are software pipelined, the number of values simultaneously alive becomes exactly known giving better opportunities for kernel loop unrolling. Furthermore, fixing circular lifetime intervals allows us to reduce the algorithmic complexity of our method compared to [1] by computing a new research space for minimal kernel loop unrolling. The meeting graph (MG) is one of the [3] frameworks proposed in the literature which models loop unrolling and register allocation together in a common formal framework for software pipelined loops. Although MG significantly improves loop register allocation, the computed loop unrolling may lead to unpractical code growth.

  12. Are Coronal Loops Isothermal or Multithermal? Yes!

    CERN Document Server

    Schmelz, J T; Rightmire, L A; Kimble, J A; Del Zanna, G; Cirtain, J W; DeLuca, E E; Mason, H E

    2009-01-01

    Surprisingly few solar coronal loops have been observed simultaneously with TRACE and SOHO/CDS, and even fewer analyses of these loops have been conducted and published. The SOHO Joint Observing Program 146 was designed in part to provide the simultaneous observations required for in-depth temperature analysis of active region loops and determine whether these loops are isothermal or multithermal. The data analyzed in this paper were taken on 2003 January 17 of AR 10250. We used TRACE filter ratios, emission measure loci, and two methods of differential emission measure analysis to examine the temperature structure of three different loops. TRACE and CDS observations agree that Loop 1 is isothermal with Log T $=$ 5.85, both along the line of sight as well as along the length of the loop leg that is visible in the CDS field of view. Loop 2 is hotter than Loop 1. It is multithermal along the line of sight, with significant emission between 6.2 $<$ Log T $<$ 6.4, but the loop apex region is out of the CDS ...

  13. Gauge theory loop operators and Liouville theory

    Energy Technology Data Exchange (ETDEWEB)

    Drukker, Nadav [Humboldt Univ. Berlin (Germany). Inst. fuer Physik; Gomis, Jaume; Okuda, Takuda [Perimeter Inst. for Theoretical Physics, Waterloo, ON (Canada); Teschner, Joerg [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany)

    2009-10-15

    We propose a correspondence between loop operators in a family of four dimensional N=2 gauge theories on S{sup 4} - including Wilson, 't Hooft and dyonic operators - and Liouville theory loop operators on a Riemann surface. This extends the beautiful relation between the partition function of these N=2 gauge theories and Liouville correlators found by Alday, Gaiotto and Tachikawa. We show that the computation of these Liouville correlators with the insertion of a Liouville loop operator reproduces Pestun's formula capturing the expectation value of a Wilson loop operator in the corresponding gauge theory. We prove that our definition of Liouville loop operators is invariant under modular transformations, which given our correspondence, implies the conjectured action of S-duality on the gauge theory loop operators. Our computations in Liouville theory make an explicit prediction for the exact expectation value of 't Hooft and dyonic loop operators in these N=2 gauge theories. The Liouville loop operators are also found to admit a simple geometric interpretation within quantum Teichmueller theory as the quantum operators representing the length of geodesics. We study the algebra of Liouville loop operators and show that it gives evidence for our proposal as well as providing definite predictions for the operator product expansion of loop operators in gauge theory. (orig.)

  14. Efficient Loop Navigation for Symbolic Execution

    CERN Document Server

    Obdrzalek, Jan

    2011-01-01

    Symbolic execution is a successful and very popular technique used in software verification and testing. A key limitation of symbolic execution is in dealing with code containing loops. The problem is that even a single loop can generate a huge number of different symbolic execution paths, corresponding to different number of loop iterations and taking various paths through the loop. We introduce a technique which, given a start location above some loops and a target location anywhere below these loops, returns a feasible path between these two locations, if such a path exists. The technique infers a collection of constraint systems from the program and uses them to steer the symbolic execution towards the target. On reaching a loop it iteratively solves the appropriate constraint system to find out which path through this loop to take, or, alternatively, whether to continue below the loop. To construct the constraint systems we express the values of variables modified in a loop as functions of the number of ...

  15. The analysis of SCS return momentum effects on the RCS water level during mid-loop operations

    Energy Technology Data Exchange (ETDEWEB)

    swang Seo, J.; Young Yang, J.; Tack Hwang, S. [Seoul National Univ. (Korea, Republic of)

    1995-09-01

    An accurate prediction of Reactor Coolant System (RCS) water levels is of importance in the determination of allowable operating range to ensure the safety during the mid-loop operations. However, complex hydraulic phenomena induced by Shutdown Cooling System (SCS) return momentum cause different water levels from those in the loop where the water level indicators are located. This was apparantly observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level and its general trend, a model using one-dimensional momentum equation, hydraulic jump, Bernoulli equation, flow resistance coefficient, and total water volume conservation has been developed to predict the RCS water levels at various RCS locations during the mid-loop conditions and the simulation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs in conjunction with the momentum loss throughout the RCS is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels.

  16. Application of CATE 2.0 code on evaluating activated corrosion products in a PWR cooling loop

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Jingyu; Li, Lu; Chen, Yixue [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering

    2017-03-15

    In PWR plants, most Occupational Radiation Exposure (ORE) for personnel results from Activated Corrosion Products (ACPs) in the cooling loop. In order to evaluate the ACPs in the cooling loop, a three-region transport model is built up based on the theory of driving force from the concentration difference in CATE 2.0 code. In order to analyze the nuclide composition of ACPs, the EAF-2007 nuclear database is embedded in CATE 2.0. The case of MIT PCCL test loop is simulated to test the availability of CATE 2.0 on PWR ACPs evaluation, and the activity of Co-58 and Co-60 after operation for 42 days calculated by CATE 2.0 is consistent with that from the code CRUDSIM adopted by MIT. Then, the nuclide composition of ACPs is analyzed in detail respectively for operation of 42 days and 12 months using CATE 2.0. The results show that the short-lived nuclides contribute a majority of the activity in the regions of in-flux wall and coolant, while the long-lived nuclides contribute most of the activity in the region of out-flux wall.

  17. Efforts to Reduce International Space Station Crew Maintenance for the Management of the Extravehicular Mobility Unit Transport Loop Water Quality

    Science.gov (United States)

    Steele, John W.; Etter, David; Rector, Tony; Boyle, Robert; Vandezande, Christopher

    2013-01-01

    The EMU (Extravehicular Mobility Unit) contains a semi-closed-loop re-circulating water circuit (Transport Loop) to absorb heat into a LCVG (Liquid Coolant and Ventilation Garment) worn by the astronaut. A second, single-pass water circuit (Feed-water Loop) provides water to a cooling device (Sublimator) containing porous plates, and that water sublimates through the porous plates to space vacuum. The cooling effect from the sublimation of this water translates to a cooling of the LCVG water that circulates through the Sublimator. The quality of the EMU Transport Loop water is maintained through the use of a water processing kit (ALCLR Airlock Cooling Loop Remediation) that is used to periodically clean and disinfect the water circuit. Opportunities to reduce crew time associated with on-orbit ALCLR operations include a detailed review of the historical water quality data for evidence to support an extension to the implementation cycle. Furthermore, an EMU returned after 2-years of use on the ISS (International Space Station) is being used as a test bed to evaluate the results of extended and repeated ALCLR implementation cycles. Finally, design, use and on-orbit location enhancements to the ALCLR kit components are being considered to allow the implementation cycle to occur in parallel with other EMU maintenance and check-out activities, and to extend the life of the ALCLR kit components. These efforts are undertaken to reduce the crew-time and logistics burdens for the EMU, while ensuring the long-term health of the EMU water circuits for a post-Shuttle 6-year service life.

  18. Higher Loop Nonplanar Anomalous Dimensions from Symmetry

    CERN Document Server

    Koch, Robert de Mello; Messamah, Ilies

    2013-01-01

    In this article we study the action of the one loop dilatation operator on operators with a classical dimension of order N. These operators belong to the su(2) sector and are constructed using two complex fields Y and Z. For these operators non-planar diagrams contribute already at the leading order in N and the planar and large N limits are distinct. The action of the one loop and the two loop dilatation operator reduces to a set of decoupled oscillators and factorizes into an action on the Z fields and an action on the Y fields. Direct computation has shown that the action on the Y fields is the same at one and two loops. In this article, using the su(2) symmetry algebra as well as structural features of field theory, we give compelling evidence that the factor in the dilatation operator that acts on the Ys is given by the one loop expression, at any loop order.

  19. Abstract Acceleration of General Linear Loops

    OpenAIRE

    2014-01-01

    International audience; We present abstract acceleration techniques for computing loop invariants for numerical programs with linear assignments and conditionals. Whereas abstract interpretation techniques typically over-approximate the set of reachable states iteratively, abstract acceleration captures the effect of the loop with a single, non-iterative transfer function applied to the initial states at the loop head. In contrast to previous acceleration techniques, our approach applies to a...

  20. Wilson Loops in Open String Theory

    Science.gov (United States)

    Shiraishi, Kiyoshi

    Wilson loop elements on torus are introduced into the partition function of open strings as Polyakov’s path integral at one-loop level. Mass spectra from compactification and expected symmetry breaking are illustrated by choosing the correct weight for the contributions from annulus and Möbius strip. We show that Jacobi’s imaginary transformation connects the mass spectra with the Wilson loops. The application to thermopartition function and cosmological implications are briefly discussed.