WorldWideScience

Sample records for lwr coolant systems

  1. 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.

  2. Evaluation of inorganic sorbent treatment for LWR coolant process streams

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, J.W.

    1984-03-01

    This report presents results of a survey of the literature and of experience at selected nuclear installations to provide information on the feasibility of replacing organic ion exchangers with inorganic sorbents at light-water-cooled nuclear power plants. Radioactive contents of the various streams in boiling water reactors and pressurized water reactors were examined. In addition, the methods and performances of current methods used for controlling water quality at these plants were evaluated. The study also includes a brief review of the physical and chemical properties of selected inorganic sorbents. Some attributes of inorganic sorbents would be useful in processing light water reactor (LWR) streams. The inorganic resins are highly resistant to damage from ionizing radiation, and their exchange capacities are generally equivalent to those of organic ion exchangers. However, they are more limited in application, and there are problems with physical integrity, especially in acidic solutions. Research is also needed in the areas of selectivity and anion removal before inorganic sorbents can be considered as replacements for the synthetic organic resins presently used in LWRs. 11 figures, 14 tables.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Regulatory instrument review: Aging management of LWR cables, containment and basemat, reactor coolant pumps, and motor-operated valves

    Energy Technology Data Exchange (ETDEWEB)

    Werry, E.V.; Somasundaram, S.

    1995-09-01

    The results of Stage 2 of the Regulatory Instrument Review are presented in this volume. Selected regulatory instruments, such as the Code of Federal Regulations (CFR), US Nuclear Regulatory Commission (NRC), Regulatory Guides, and ASME Codes, were investigated to determine the extent to which these regulations apply aging management to selected safety-related components in nuclear power plants. The Regulatory Instrument Review was funded by the NRC under the Nuclear Plant Aging Research (NPAR) program. Stage 2 of the review focused on four safety-related structures and components; namely, cables, containment and basemat, reactor coolant pumps, and motor-operated valves. The review suggests that the primary-emphasis of the regulatory instruments was on the design, construction, start-up, and operation of a nuclear power plant, and that aging issues were primarily addressed after an aging-related problem was recognized. This Stage 2 review confirms the results of the prior review; (see Regulatory Instrument Review: Management of Aging of LWR Major Safety-Related Components NUREG/CR-5490. The observations indicate that the regulations generally address management of age-related degradation indirectly. Specific age-related degradation phenomena frequently are dealt with in bulletins and notices or through generic issues, letters, etc. The major recommendation of this report, therefore, is that the regulatory instruments should more directly and explicitly address the aging phenomenon and the management of the age-related degradation process.

  8. Development of information management system on LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility.

  9. 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

  10. LWR and defectivity improvement on EUV track system

    Science.gov (United States)

    Harumoto, Masahiko; Stokes, Harold; Thouroude, Yan; Kaneyama, Koji; Pieczulewski, Charles; Asai, Masaya

    2016-03-01

    EUV lithography (EUVL) is well known to be a strong candidate for next generation, single exposure sub-30nm halfpitch lithography.[1] Furthermore, high-NA EUV exposure tool(s) released two years ago gave a strong impression by finer pattern results. On the other hand, it seems that the coat-develop track process remains very similar and in many aspects returns to KrF or ArF dry process fundamentals, but in practice a 26-32nm pitch patterning coat develop track process also has challenges with EUV resists. As access to EUV lithography exposures has become more readily available over the last five (5) years, several challenges and accomplishments in the track process have been reported, such as the improvement of ultra-thin film coating, CD uniformity, defectivity, line width roughness (LWR), and so on.[2-8] The coat-develop track process has evolved along with novel materials and metrology capability. Line width roughness (LWR) control and defect reduction are demonstrated utilizing the SOKUDO DUO coat-develop track system with ASML NXE:3100 and NXE:3300 exposures in the IMEC (Leuven, Belgium) cleanroom environment. Additionally, we will show the latest lithographic results obtained by novel processing approaches in the EUV coat develop track system.

  11. Boron concentration control system for LWR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fukae, Chiyokazu (Kansai Electric Power Co., Inc., Osaka (Japan)); Tanimoto, Kazuo; Suzuki, Hideaki; Ishiguro, Akira; Yoshiki, Ryoji; Nakakura, Hiroyuki.

    1993-04-23

    The present invention mitigates a delay of control response upon concentration and reverse operation upon dilution of boron in coolants, to ensure good control response and system stability. That is, in an existent boron concentration control system, an injection system of aqueous boric acid at predetermined concentration and a dilution water injection system are connected to a coolant recycling system by way of a mixer. An adder for receiving axial distribution parameter, and a set value thereof and a calculator for outputting a dilution water flow rate demand signal or an aqueous boric acid flow rate demand signal upon receiving the output from the adder are disposed to the system. The aqueous boric acid flow rate demand signal is applied to an aqueous boric acid injection flow rate control valve, as well as dilution water flow rate demand signal upon concentration is applied to a dilution water flow rate control valve. With such procedures, the delay for the transportation of aqueous boric acid caused when the boric acid in primary coolant system is changed from dilution to concentration is improved. On the other hand, the reverse operation of temporal concentration caused when the concentration is changed to dilution is decreased. Further, stability of control system is also sufficient. (I.S.).

  12. 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.

  13. 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. 

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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)

  19. LWR fuel rod behavior during reactor tests under loss-of-coolant conditions: Results of the FR2 in-pile tests

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.; Sepold, L.; Hofmann, P.; Petersen, C.; Schanz, G.; Zimmermann, H. (Kernforschungszentrum Karlsruhe G.m.b.H. (Germany, F.R.))

    1982-05-01

    Results of the FR2 in-pile tests on fuel rod behavior under loss-of-coolant accident (LOCA) conditions are presented. To investigate the possible influence of a nuclear environment on fuel rod failure mechanisms, unirradiated as well as irradiated (2500 to 35,000 MWd/tsub(U)) PWR-type test fuel rods were exposed to temperature transients simulating the second heatup phase of a LOCA. Loaded by internal overpressure, the cladding ballooned and ruptured. The burst data do not indicate major differences from results obtained out-of-pile with electrically heated fuel rod simulators, and do not show an influence of burnup. The fuel pellets in previously irradiated rods, already cracked during normal operation, crumbled completely in the regions with large cladding deformation. Post-test examinations revealed cladding mechanical behavior and oxidation to be comparable to out-of-pile results, with relatively little fission gas release during the transient.

  20. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  1. 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.

  2. Status report on assessment of environmentally assisted fatigue for LWR extended service conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, W. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-07-09

    This report provides an update on an earlier assessment of environmentally assisted fatigue for light water reactor (LWR) materials under extended service conditions. This report is a deliverable in September 2013, under the work package for environmentally assisted fatigue in the Light Water Reactor Sustainability (LWRS) program. The overall objective of this LWRS project is to assess the degradation by environmentally assisted cracking/fatigue of LWR materials, such as various alloy base metals and their welds used in reactor coolant system piping. This effort is to support the U.S. Department of Energy LWRS program for developing tools to predict the aging/failure mechanism and to correspondingly predict the remaining life of LWR components for anticipated 60-80 year operation.

  3. 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.

  4. 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.

  5. Ice Thermal Storage Systems for LWR Supplemental Cooling and Peak Power Shifting

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Phil Sharpe; Blaise Hamanaka; Wei Yan; WoonSeong Jeong

    2010-06-01

    Availability of enough cooling water has been one of the major issues for the nuclear power plant site selection. Cooling water issues have frequently disrupted the normal operation at some nuclear power plants during heat waves and long draught. The issues become more severe due to the new round of nuclear power expansion and global warming. During hot summer days, cooling water leaving a power plant may become too hot to threaten aquatic life so that environmental regulations may force the plant to reduce power output or even temporarily to be shutdown. For new nuclear power plants to be built at areas without enough cooling water, dry cooling can be used to remove waste heat directly into the atmosphere. However, dry cooling will result in much lower thermal efficiency when the weather is hot. One potential solution for the above mentioned issues is to use ice thermal storage systems (ITS) that reduce cooling water requirements and boost the plant’s thermal efficiency in hot hours. ITS uses cheap off-peak electricity to make ice and uses those ice for supplemental cooling during peak demand time. ITS is suitable for supplemental cooling storage due to its very high energy storage density. ITS also provides a way to shift large amount of electricity from off peak time to peak time. Some gas turbine plants already use ITS to increase thermal efficiency during peak hours in summer. ITSs have also been widely used for building cooling to save energy cost. Among three cooling methods for LWR applications: once-through, wet cooling tower, and dry cooling tower, once-through cooling plants near a large water body like an ocean or a large lake and wet cooling plants can maintain the designed turbine backpressure (or condensation temperature) during 99% of the time; therefore, adding ITS to those plants will not generate large benefits. For once-through cooling plants near a limited water body like a river or a small lake, adding ITS can bring significant economic

  6. 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.

  7. 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.

  8. 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.

  9. 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)

  10. 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.

  11. Chemical thermodynamics of complex systems: fission product behavior in LWR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Kohli, R.

    1981-03-01

    A detailed thermodynamic assessment has been made of the chemical reactions of fission products in LWR fuel rods. Using recent thermodynamic data and the in-reactor oxygen potential and temperature range of LWRs, equilibrium thermodynamic calculations were performed for the most plausible reactions of the fission products. The emphasis in this model is on the chemistry of cesium and rubidium and their reactions with the fuel, other fission products, and the zircaloy cladding. The model predictions are discussed for their implications in fuel-cladding interactions.

  12. A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

    Directory of Open Access Journals (Sweden)

    Jorge Pérez Mañes

    2014-01-01

    Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

  13. 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.

  14. 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.

  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. 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%.

  17. Cold atoms as a coolant for levitated optomechanical systems

    CERN Document Server

    Ranjit, Gambhir; Geraci, Andrew A

    2014-01-01

    Optically trapped dielectric objects are well suited for reaching the quantum regime of their center of mass motion in an ultra-high vacuum environment. We show that ground state cooling of an optically trapped nanosphere is achievable when starting at room temperature, by sympathetic cooling of a cold atomic gas optically coupled to the nanoparticle. Unlike cavity cooling in the resolved sideband limit, this system requires only a modest cavity finesse and it allows the cooling to be turned off, permitting subsequent observation of strongly-coupled dynamics between the atoms and sphere. Nanospheres cooled to their quantum ground state could have applications in quantum information science or in precision sensing.

  18. 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)

  19. 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.)

  20. 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.

  1. Assessment of LWR-HTR-GCFR Integrated Cycle

    Directory of Open Access Journals (Sweden)

    Eleonora Bomboni

    2009-01-01

    Full Text Available Preliminary analyses already performed showed that innovative GCRs, both thermal and fast, are very promising candidate to reach the Gen-IV sustainability goal. The integrated LWR-HTR-GCFR basically aims at closing the current nuclear fuel cycle: in principle, thanks to the unique characteristics of Helium coolant reactors, LWR SNF along with DU become valuable material to produce energy. Additionally, burning HMs of LWR SNF means not only a drastic reduction in the Unat demand but also a remarkable decrease in the long-term radiotoxic component of nuclear waste to be geologically stored. This paper focuses on the analyses of the LWR-HTR-GCFR cycle performed by the University of Pisa in the frame of the EU PUMA project (6th FP. Starting from a brief outline of the main characteristics of HTR and GCFR concepts and of the advantages of linking LWR, HTR and GCFR in a symbiotic way, this paper shows the integrated cycle involving a typical LWR (1000 MWe, a PBMR (400 MWth and a GCFR-“E” (2400 MWth. Additionally, a brief overview of the main technological constraints concerning (Pu+MA-based advanced fuels is given, in order to explain and justify the choices made in the framework of the considered cycle. Thereafter, calculations performed and results obtained are described.

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. LIFE vs. LWR: End of the Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J C; Blink, J A; Shaw, H F

    2008-10-02

    The worldwide energy consumption in 2003 was 421 quadrillion Btu (Quads), and included 162 quads for oil, 99 quads for natural gas, 100 quads for coal, 27 quads for nuclear energy, and 33 quads for renewable sources. The projected worldwide energy consumption for 2030 is 722 quads, corresponding to an increase of 71% over the consumption in 2003. The projected consumption for 2030 includes 239 quads for oil, 190 quads for natural gas, 196 quads for coal, 35 quads for nuclear energy, and 62 quads for renewable sources [International Energy Outlook, DOE/EIA-0484, Table D1 (2006) p. 133]. The current fleet of light water reactors (LRWs) provides about 20% of current U.S. electricity, and about 16% of current world electricity. The demand for electricity is expected to grow steeply in this century, as the developing world increases its standard of living. With the increasing price for oil and gasoline within the United States, as well as fear that our CO2 production may be driving intolerable global warming, there is growing pressure to move away from oil, natural gas, and coal towards nuclear energy. Although there is a clear need for nuclear energy, issues facing waste disposal have not been adequately dealt with, either domestically or internationally. Better technological approaches, with better public acceptance, are needed. Nuclear power has been criticized on both safety and waste disposal bases. The safety issues are based on the potential for plant damage and environmental effects due to either nuclear criticality excursions or loss of cooling. Redundant safety systems are used to reduce the probability and consequences of these risks for LWRs. LIFE engines are inherently subcritical, reducing the need for systems to control the fission reactivity. LIFE engines also have a fuel type that tolerates much higher temperatures than LWR fuel, and has two safety systems to remove decay heat in the event of loss of coolant or loss of coolant flow. These features of

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations - Burnup Analysis

    Science.gov (United States)

    Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

    2014-06-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

  12. 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)

  13. 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).

  14. 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.

  15. 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.

  16. 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

  17. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  18. 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.

  19. 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%.

  20. 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

  1. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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)

  7. Long-term embrittlement of cast duplex stainless steels in LWR systems. Semiannual report, October 1991--March 1992: Volume 7, No.1

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K. [Argonne National Lab., IL (United States)

    1993-05-01

    This progress report summarizes work performed by Argonne National Laboratory on long-term thermal embrittlement of cast duplex stainless steels in LWR systems during the six months from October 1991 to March 1992. Charpy-impact, tensile, and fracture toughness J-R curve data are presented for several heats of cast stainless steel that were aged 10,000-58,000 h at 290, 320, and 350{degree}C. The results indicate that thermal aging decreases the fracture toughness of cast stainless steels. In general, CF-3 steels are the least sensitive to thermal aging and CF-8M steels are the most sensitive. The values of fracture toughness J{sub IC} and tearing modulus for CF-8M steels can be as low as {approx}90 kJ/m{sup 2} and {approx}60, respectively. The fracture toughness data are consistent with the Charpy-impact results, i.e. unaged and aged steels that show low impact energy also exhibit lower fracture toughness. All steels reach a minimum saturation fracture toughness after thermal aging; the time to reach saturation depends on the aging temperature. The results also indicate that low-strength cast stainless steels are generally insensitive to thermal aging.

  8. Survey of tracking systems and rotary joints for coolant piping. Final report, August 15, 1978-August 14, 1978. [Includes patents

    Energy Technology Data Exchange (ETDEWEB)

    Furaus, J P; Gruchalla, M E; Sower, G D

    1980-01-01

    Problems were surveyed and evaluated with respect to solar tracking mechanisms and rotary joints for coolant piping. An analytical development of celestial mechanics, one- and two-axis tracking configurations and the effect of tracking accuracy versus collector efficiency are reported. Daily operational requirements and tracking modes were defined and evaluated. A literature and patent search on solar tracking technology was performed. Tracking system and control system performance specifications were determined. Alternative conceptual tracking approaches were defined and a cost and performance evaluation of a mechanical tracking concept was performed. Fluid coupling service specifications were determined. The cost and performance of several types of actuators and error detectors were evaluated with respect to solar tracking mechanisms.

  9. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in 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.

  10. 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.

  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)

    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.

  13. KSTAR Severe Accident Analysis using MELCOR : Ex-vessel Coolant Pipe Break with Failure of Fusion Power Termination System

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    To investigate the consequence of severe accidents in fusion reactor, a number of thermal hydraulics simulation codes were used (ECART, INTRA, ATHENA/RELAP and so on). MELCOR is chosen as the thermal hydraulics code to simulate the consequence of radioactive material release from accident in preliminary safety report. Capability of the simulation code for fusion reactor severe accident analysis is ability to simulate the hydraulic system in ITER and the transport phenomenon of radionuclides. MELCOR is a fully integrated code that models the accidents in Light Water Reactor (LWR). There are three kinds of radioactive materials in fusion reactor; tritium (or Tiritiated water: HTO), activation products (AP) of divertor or first-wall and activated corrosion products(ACP). In generic Site Safety Report (GSSR), the release guidelines for tritium and activation products are listed for normal operation, incidents, and accidents. And this guidelines presented in Table 1. Not only ITER, the KSTAR (Korea Superconducting Tokamak Advanced Research) is also developing fusion research reactor. The scale of facility is smaller than ITER but this small scale of facility offers the experimental flexibility to develop fusion technology. The major differences between KSTAR and ITER systems are presented in Table 2. Fusion source difference between KSTAR and ITER is D-D fusion reaction (Deuterium-Deuterium fusion reaction) and D-T fusion reaction (Deuterium-Tritium fusion reaction). This D-D fusion makes one tritium by 50 percent chance. The radioactivity of tritium is small to consider compared to radioactive materials in nuclear fission reactor. This reaction is presented in equation (1) In the present work, conservatively estimated tritium inventory amount in KSTAR is used with one of the most severe accident in ITER; Ex-vessel pipe break with Fusion Power Termination System (FPTS). The MELCOR KSTAR input is made by scaling down the ITER input deck. So, the detail system is not same

  14. 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

  15. ERDA LWR plant technology program: role of government/industry in improving LWR performance

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-07

    Information is presented under the following chapter headings: executive summary; LWR plant outages; LWR plant construction delays and cancellations; programs addressing plant outages, construction delays, and cancellations; need for additional programs to remedy continuing problems; criteria for government role in LWR commercialization; and the proposed government program.

  16. Providing the Basis for Innovative Improvements in Advanced LWR Reactor Passive Safety Systems Design: An Educational R&D Project

    Energy Technology Data Exchange (ETDEWEB)

    Brian G. Williams; Jim C. P. Liou; Hiral Kadakia; Bill Phoenix; Richard R. Schultz

    2007-02-27

    This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems.

  17. Initial guidance on digraph-matrix analysis for systems interaction studies at selected LWR's

    Energy Technology Data Exchange (ETDEWEB)

    Alesso, H.P.; Sacks, I.J.; Smith, C.F.

    1982-08-12

    The Nuclear Regulatory Commission (NRC) is currently developing guidance to search for and evaluate systems interactions at light water reactors (LWRs). In addition, related efforts are being made to incorporate such analyses into the framework of Probabilistic Risk Assessment (PRA). Preliminary results indicate that there are at least three concepts on how systems interactions can be analyzed independently or incorporated into an overall PRA. One point of view is that systems interactions can be adequately analyzed by enhancing existing PRAs. A second and closely related point of view is that systems interactions can be incorporated into a PRA in the event-tree stage of analysis. This approach attempts to capture systems interactions at an earlier stage of analysis. By treating dependencies in the event tree analysis portion of a PRA, the requirement of fault tree modeling at additional levels of detail is reduced. A third point of view about the Systems Interaction problem is that a new technique, called Digraph-Matrix Analysis (DMA), which utilizes matrix representation of logic diagrams, may offer a more complete and possibly more efficient analysis in certain areas. This report will present proposed initial guidance on DMA for application to systems interactions. DMA will be presented as a procedure independent from PRA. Peer review and later sample applications can provide useful feedback for refining the final guidance.

  18. Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

    2016-02-15

    Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

  19. Assessment of LWR spent fuel disposal options. Volume 3. Study bases and system design considerations (Appendices). Technical report

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-01

    Volume 3 (Appendices) provides a tabulation of the bases and assumptions used in the study as well as preconceptual design description and cost estimates of the facilities and transportation systems necessary to implement the various study cases.

  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. 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.

  2. Thermal modeling in an engine cooling system to control coolant flow for fuel consumption improvement

    Science.gov (United States)

    Park, Sangki; Woo, Seungchul; Kim, Minho; Lee, Kihyung

    2016-09-01

    The design and evaluation of engine cooling and lubrication systems is generally based on real vehicle tests. Our goal here was to establish an engine heat balance model based on mathematical and interpretive analysis of each element of a passenger diesel engine cooling system using a 1-D numerical model. The purpose of this model is to determine ways of optimizing the cooling and lubrication components of an engine and then to apply these methods to actual cooling and lubrication systems of engines that will be developed in the future. Our model was operated under the New European Driving Cycle (NEDC) mode conditions, which represent the fuel economy evaluation mode in Europe. The flow rate of the cooling system was controlled using a control valve. Our results showed that the fuel efficiency was improved by as much as 1.23 %, cooling loss by 1.35 %, and friction loss by 2.21 % throughout NEDC modes by modification of control conditions.

  3. Bypass valve and coolant flow controls for optimum temperatures in waste heat recovery systems

    Science.gov (United States)

    Meisner, Gregory P

    2013-10-08

    Implementing an optimized waste heat recovery system includes calculating a temperature and a rate of change in temperature of a heat exchanger of a waste heat recovery system, and predicting a temperature and a rate of change in temperature of a material flowing through a channel of the waste heat recovery system. Upon determining the rate of change in the temperature of the material is predicted to be higher than the rate of change in the temperature of the heat exchanger, the optimized waste heat recovery system calculates a valve position and timing for the channel that is configurable for achieving a rate of material flow that is determined to produce and maintain a defined threshold temperature of the heat exchanger, and actuates the valve according to the calculated valve position and calculated timing.

  4. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  5. Improving the safety of LWR power plants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-01

    This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs (improving or maintaining level of safety with simpler systems or in a more cost-effective manner).

  6. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2B, User's guide to the LWR assemblies data base, Appendix 2C, User's guide to the LWR radiological data base, Appendix 2D, User's guide to the LWR quantities data base

    Energy Technology Data Exchange (ETDEWEB)

    None

    1987-12-01

    This User's Guide for the LWR Assemblies data base system is part of the Characteristics Data Base being developed under the Waste Systems Data Development Program. The objective of the LWR Assemblies data base is to provide access at the personal computer level to information about fuel assemblies used in light-water reactors. The information available is physical descriptions of intact fuel assemblies and radiological descriptions of spent fuel disassembly hardware. The LWR Assemblies data base is a user-oriented menu driven system. Each menu is instructive about its use. Section 5 of this guide provides a sample session with the data base to assist the user.

  7. 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.

  8. 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.

  9. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  10. 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.

  11. A system for automatic biocide treatment of coolant cycles with concentration; Ein System einer automatisierten Biozidbehandlung von Kuehlkreislaeufen mit Eindickung

    Energy Technology Data Exchange (ETDEWEB)

    Lutat, A.; Pflug, H.D.; Schoenfelder, T. [KNG, Rostock (Germany)

    2000-07-01

    Regular, discontinuous treatment of industrial-scale recirculation cooling systems with microcide solutions must be carried out in view of the following aspects: Microbial growth inside the whole cooling systems during the interval times must be weakened so that no operational disturbances will occur and no additional measures for removal of organic depositions will be required. - Limiting values for microcide concentrations in the effluents must be observed. - Microcide treatment must be possible without load reduction of the power plant. - Consumption of microcide must be minimised. This means that both the time for treatment and the microcide concetration are limited. The contribution describes the automatic shock chlorination process employed at Rostock power station since a few years, i.e. water treatment and measuring instruments, investigation of microbial growth in the cooling water of a pilot plant, concept of biocide treatment of Rostock power station, cooling tower operation and chlorine concentrations during chlorination. [German] Bei der regelmaessigen, diskontinuierlichen Behandlung grosser Rueckkuehlsysteme mit Mikroziden, im allgemeinen chloraktiven Loesungen, sind mehrere Forderungen zu beruecksichtigen, naemlich: - Das mikrobielle Wachstum im gesamten Kuehlsystem muss in den Intervallzeiten soweit geschwaecht sein, dass es zu keinen betrieblichen Stoerungen kommt und moeglichst auch keine zusaetzlichen Massnahmen zur Entfernung von organischem Bewuchs noetig werden. - Die behoerdlichen Auflagen bezueglich des tolerierbaren Restgehalts an Mikrozid in der Abflut sind einzuhalten. - Eine Mikrozidbehandlung sollte ohne Lasteinschraenkung des Kraftwerks durchgefuehrt werden koennen. - Der Verbrauch an Mikrozid ist zu minimieren. Das bedeutet, dass nur eine begrenzte Betriebszeit bei geschlossener Kuehlturmabflut zur Verfuegung steht, um die wirksame Konzentration an Mikrozid bis zu dem behoerdlich zugelassenen Grenzwert abzubauen. Eine Ueberdosis

  12. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    Energy Technology Data Exchange (ETDEWEB)

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment.

  13. Phased Array Ultrasonic Examination of Reactor Coolant System (Carbon Steel-to-CASS) Dissimilar Metal Weld Mockup Specimen

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, S. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cinson, A. D. [US Nuclear Regulatory Commission (NRC), Washington, DC (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anderson, M. T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-11-23

    In the summer of 2009, Pacific Northwest National Laboratory (PNNL) staff traveled to the Electric Power Research Institute (EPRI) NDE Center in Charlotte, North Carolina, to conduct phased-array ultrasonic testing on a large bore, reactor coolant pump nozzle-to-safe-end mockup. This mockup was fabricated by FlawTech, Inc. and the configuration originated from the Port St. Lucie nuclear power plant. These plants are Combustion Engineering-designed reactors. This mockup consists of a carbon steel elbow with stainless steel cladding joined to a cast austenitic stainless steel (CASS) safe-end with a dissimilar metal weld and is owned by Florida Power & Light. The objective of this study, and the data acquisition exercise held at the EPRI NDE Center, were focused on evaluating the capabilities of advanced, low-frequency phased-array ultrasonic testing (PA-UT) examination techniques for detection and characterization of implanted circumferential flaws and machined reflectors in a thick-section CASS dissimilar metal weld component. This work was limited to PA-UT assessments using 500 kHz and 800 kHz probes on circumferential flaws only, and evaluated detection and characterization of these flaws and machined reflectors from the CASS safe-end side only. All data were obtained using spatially encoded, manual scanning techniques. The effects of such factors as line-scan versus raster-scan examination approaches were evaluated, and PA-UT detection and characterization performance as a function of inspection frequency/wavelength, were also assessed. A comparative assessment of the data is provided, using length-sizing root-mean-square-error and position/localization results (flaw start/stop information) as the key criteria for flaw characterization performance. In addition, flaw signal-to-noise ratio was identified as the key criterion for detection performance.

  14. Iodine chemical forms in LWR severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Beahm, E.C.; Weber, C.F.; Kress, T.S.; Parker, G.W.

    1991-01-01

    Calculated data from seven severe accident sequences in light-water reactor plants were used to assess the chemical forms of iodine in containment. In most of the calculations for the seven sequences, iodine entering containment from the reactor coolant system was almost entirely in the form of CsI with very small contributions of I or HI. The largest fraction of iodine in forms other than CsI was a total of 3.2% as I plus HI. Within the containment, the CsI will deposit onto walls and other surfaces, as well as in water pools, largely in the form of iodide (I{sup {minus}}). The radiation induced conversion of I{sup {minus}} in water pools into I{sub 2} is strongly dependent on pH. In systems where the pH was controlled above 7, little additional elemental iodine would be produced in the containment atmosphere. When the pH falls below 7, it may be assumed that it is not being controlled, and large fractions of iodine as I{sub 2} within the containment atmosphere may be produced. 16 refs.

  15. Development of tool for implementation of energy economic cooling systems with natural coolants. Part 1: Installation and calculation of annual energy consumption; Udvikling af vaerktoej til implementering af energioekonomiske koeleanlaeg med naturlige koelemidler. Del 1: Opstilling og beregning af aersenergiforbrug

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-02-15

    This report presents the results reached during the project 'Development of tools for implementation of energy economic cooling systems with natural coolants. Part 1: Installation and calculation of annual energy consumption'. The report is primarily directed at grant-awarding authorities, as the results only propose recommendations concerning further research. The aim of making a tool is, quite simply, to compare the economy for alternative designs of cooling systems when both investments and operational costs during the system's service life are taken into consideration. Thus the tool gives a buyer of a cooling system the possibility of choosing the most energy economic system, and hereby spread the most energy economic systems in connection with the system replacement that will take place due to out phasing of CFC/HCFC/HFC coolants. The project has been divided into two phases. Phase no. 1, this phase, contains the development of the tool. (BA)

  16. 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.

  17. LWR spent fuel reduction by the removal of U and the compact storage of Pu with FP for long-term nuclear sustainability

    Energy Technology Data Exchange (ETDEWEB)

    Fukasawa, T.; Hoshino, K. [Hitachi-GE Nuclear Energy, Ltd, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Takano, M. [Japan Atomic Energy Agency, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Sato, S. [Hokkaido University, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Shimazu, Y. [Fukui University, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan)

    2013-07-01

    Fast breeder reactors (FBR) nuclear fuel cycle is needed for long-term nuclear sustainability while preventing global warming and maximum utilizing the limited uranium (U) resources. The 'Framework for Nuclear Energy Policy' by the Japanese government on October 2005 stated that commercial FBR deployment will start around 2050 under its suitable conditions by the successive replacement of light water reactors (LWR) to FBR. Even after Fukushima Daiichi Nuclear Power Plant accident which made Japanese tendency slow down the nuclear power generation activities, Japan should have various options for energy resources including nuclear, and also consider the delay of FBR deployment and increase of LWR spent fuel (LWR-SF) storage amounts. As plutonium (Pu) for FBR deployment will be supplied from LWR-SF reprocessing and Japan will not possess surplus Pu, the authors have developed the flexible fuel cycle initiative (FFCI) for the transition from LWR to FBR. The FFCI system is based on the possibility to stored recycled materials (U, Pu)temporarily for a suitable period according to the FBR deployment rate to control the Pu demand/supply balance. This FFCI system is also effective after the Fukushima accident for the reduction of LWR-SF and future LWR-to-FBR transition. (authors)

  18. Development of FUELSIM/MOD0 for the detailed analysis of LWR fuel rod behavior under normal operation conditions with extended burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G.A.; Allison, C.M. [Innovative Systems Software LLC, 1284 South Woodruff, Idaho Falls, ID (United States)

    1999-07-01

    The FUELSIM code is being developed by Innovative Systems Software as part of the international SCDAP Development and Training Program. FUELSIM is being developed as a 'stand-alone' best-estimate fuel behavior code with evaluation modeling options. The long term goal of the code is to predict fuel performance over the full range of conditions from normal operating behavior to severe accident conditions using a combination of models from the FRAPCON-3, FRAP-T6, SCDAP, and MATPRO fuel behavior codes. FUELSIM/MOD0 is the first release of the code and includes models to predict the behavior of LWR fuel rods during normal operating conditions including the influence of extended burnup fuel. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The code models all the important phenomena that occur during normal operating conditions and contains necessary materials properties, water properties, and heat transfer correlations. The code runs on a variety of computers and operating systems including UNIX, LINUX, and Windows NT or 95. (author)

  19. Multilevel transport solution of LWR reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Cassiano R.E. de Oliveira; HyeonKae Park

    2008-09-01

    This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations.

  20. Development and application of an information-analytic system on the problem of flow accelerated corrosion of pipeline elements in the secondary coolant circuit of VVER-440-based power units at the Novovoronezh nuclear power plant

    Science.gov (United States)

    Tomarov, G. V.; Povarov, V. P.; Shipkov, A. A.; Gromov, A. F.; Kiselev, A. N.; Shepelev, S. V.; Galanin, A. V.

    2015-02-01

    Specific features relating to development of the information-analytical system on the problem of flow-accelerated corrosion of pipeline elements in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh nuclear power plant are considered. The results from a statistical analysis of data on the quantity, location, and operating conditions of the elements and preinserted segments of pipelines used in the condensate-feedwater and wet steam paths are presented. The principles of preparing and using the information-analytical system for determining the lifetime to reaching inadmissible wall thinning in elements of pipelines used in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh NPP are considered.

  1. Tornado wind-loading requirements based on risk assessment techniques (For specific reactor safety Class 1 coolant system features)

    Science.gov (United States)

    Deobald, Theodore L.; Coles, Garill A.; Smith, Gary L.

    1992-01-01

    Regulations require that nuclear power plants be protected from tornado winds. If struck by a tornado, a plant must be capable of safely shutting down and removing decay heat. Probabilistic techniques are used to show that risk to the public from the U.S. Department of Energy SP-100 reactor is acceptable without tornado hardening parts of the secondary system. Relaxed requirements for design wind loadings will result in significant cost savings. To demonstrate an acceptable level of risk, this document examines tornado-initiated accidents. The two tornado-initiated accidents examined in detail are loss of cooling resulting in core damage and loss of secondary system boundary integrity leading to sodium release. Loss of core cooling is analyzed using fault/event tree models. Loss of secondary system boundary integrity is analyzed by comparing the consequences to acceptance criteria for the release of radioactive material or alkali metal aerosol.

  2. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 2: with Beam Shutdown Only

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report. This report documents the results of simulations of a Loss-of-Flow Accident (LOFA) where power is lost to all of the pumps that circulate water in the blanket region, the accelerator beam is shut off and neither the residual heat removal nor cavity flood systems operate.

  3. ANALYSIS OF THE IMPACT PROPERTIES OF THE COOLANT RECOVERY SYSTEM HEAT LOSSES OF COMBINED COMPRESSOR-POWER PLANT ON ITS CHARACTERISTICS

    Directory of Open Access Journals (Sweden)

    Yusha V.L.

    2012-12-01

    Full Text Available The paper presents results of theoretical analysis of the effectiveness of an ideal thermodynamic cycle internal combustion engine combined with an external utilization of exhaust heat. The influence of the properties of the coolant circuit of utilization on its operational parameters and characteristics of the power plant.

  4. Parameters influencing the transgranular stress corrosion cracking behaviour of austenitic stainless steels in systems conveying reactor coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kilian, R.; Wesseling, U. [Framatome ANP (Germany); Wachter, O. [E.ON Kernkraft (Germany); Widera, M. [RWE Power (Germany); Brummer, G. [HEW - (Germany); Ilg, U. [EnBW - (Germany)

    2002-07-01

    During replacement of an auxiliary system in the German PWR KKS (NPP Stade) a damage was detected in a valve housing and in the connected piping both made from stabilised austenitic stainless steel. During operation stagnant conditions are present in this area. Based on the failure analysis chloride induced stress corrosion cracking (SCC) was found as the dominating root cause. In the open literature many cases of corrosion observed in the water/steam interface in valve components as well as in adjacent portions of auxiliary circuits made of un-stabilized stainless steels are mentioned. A common feature of the reported cases is that transgranular cracking was found. Extensive laboratory investigations revealed that non-stabilised austenitic stainless steels are also sensitive to transgranular cracking in boric acid solutions particularly in concentrated solutions. Often these solutions are contaminated with chlorides and/or oxygen is present. Taking into account the literature data the question could arise whether the above mentioned cracking may be also caused by boric acid attack. Thus, for stabilised stainless steels laboratory exposure tests at 80 C in saturated aerated boric acid solution and at 300 C in (at 100 C) saturated, oxygen free boric acid solution have been performed. Double-U-bend specimens and wedge loaded 1T-CT specimens made of Ti- and Nb-stabilised austenitic stainless steels were used. The results revealed no evidence of crack initiation and crack growth. Based on the laboratory results and the literature data an attempt is undertaken to separate parameters influencing chloride induced SCC from the effect of boric acid. (authors)

  5. ELCOS: the PSI code system for LWR core analysis. Part II: user`s manual for the fuel assembly code BOXER

    Energy Technology Data Exchange (ETDEWEB)

    Paratte, J.M.; Grimm, P.; Hollard, J.M. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-02-01

    ELCOS is a flexible code system for the stationary simulation of light water reactor cores. It consists of the four computer codes ETOBOX, BOXER, CORCOD and SILWER. The user`s manual of the second one is presented here. BOXER calculates the neutronics in cartesian geometry. The code can roughly be divided into four stages: - organisation: choice of the modules, file manipulations, reading and checking of input data, - fine group fluxes and condensation: one-dimensional calculation of fluxes and computation of the group constants of homogeneous materials and cells, - two-dimensional calculations: geometrically detailed simulation of the configuration in few energy groups, - burnup: evolution of the nuclide densities as a function of time. This manual shows all input commands which can be used while running the different modules of BOXER. (author) figs., tabs., refs.

  6. 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.

  7. Options for Burning LWR SNF in LIFE Engine

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-09

    We have pursued two processes in parallel for the burning of LWR SNF in the LIFE engine: (1) solid fuel option and (2) liquid fuel option. Approaches with both are discussed. The assigned Topical Report on liquid fuels is attached.

  8. 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)

  9. Data center coolant switch

    Energy Technology Data Exchange (ETDEWEB)

    Iyengar, Madhusudan K.; Parida, Pritish R.; Schultz, Mark D.

    2015-10-06

    A data center cooling system is operated in a first mode; it has an indoor portion wherein heat is absorbed from components in the data center, and an outdoor heat exchanger portion wherein outside air is used to cool a first heat transfer fluid (e.g., water) present in at least the outdoor heat exchanger portion of the cooling system during the first mode. The first heat transfer fluid is a relatively high performance heat transfer fluid (as compared to the second fluid), and has a first heat transfer fluid freezing point. A determination is made that an appropriate time has been reached to switch from the first mode to a second mode. Based on this determination, the outdoor heat exchanger portion of the data cooling system is switched to a second heat transfer fluid, which is a relatively low performance heat transfer fluid, as compared to the first heat transfer fluid. It has a second heat transfer fluid freezing point lower than the first heat transfer fluid freezing point, and the second heat transfer fluid freezing point is sufficiently low to operate without freezing when the outdoor air temperature drops below a first predetermined relationship with the first heat transfer fluid freezing point.

  10. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  11. A Stochastic LWR Model with Consideration of the Driver's Individual Property%A Stochastic LWR Model with Consideration of the Driver's Individual Property

    Institute of Scientific and Technical Information of China (English)

    唐铁桥; 王云鹏; 余贵珍; 黄海军

    2012-01-01

    In this paper,we develop a stochastic LWR model based on the influences of the driver's individual property on his/her perceived density and speed deviation.The numerical results show that the driver's individual property has great effects on traffic flow only when the initial density is moderate,i.e.,at this time,oscillating traffic flow will occur and the oscillating phenomena in the traffic system consisting of the conservative and aggressive drivers is more serious than that in the traffic system consisting of the conservative (aggressive) drivers.

  12. Micro reactor physics of MOX fueled LWR

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan)

    2001-09-01

    Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)

  13. Micro reactor physics of MOX fueled LWR

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan)

    2001-09-01

    Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)

  14. Coolants for a closed-cycle cooling system of device for removing the condensation heat of teh waste steam from steam-turbine power plants

    Energy Technology Data Exchange (ETDEWEB)

    Richter, D.; Heeren, H.

    1976-03-25

    A series of demands are made on the coolant for the purpose mentioned. They should (a) have a large heat of evaporation, a low freezing point, a high vapor density at evaporation-condensation temperature, high heat transfer numbers with bubble vapors and condensation and a low saturated vapor pressure at operational temperature, (b) not be poisonous and inflammable, and (c) be economical compared to forced-water circulation. According to the invention, an azeotropic mixture of water and methyl-propylene glycol fulfills these requirements.

  15. Thermal conductivity of heterogeneous LWR MOX fuels

    Science.gov (United States)

    Staicu, D.; Barker, M.

    2013-11-01

    It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

  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. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  3. 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.

  4. 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.

  5. 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.

  6. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and

  7. 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.

  8. Survey of LWR environmental control technology performance and cost

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Aaberg, R.L.; Cole, B.M.; Engel, R.L.; Kennedy, W.E. Jr.; Lewallen, M.A.

    1980-03-01

    This study attempts to establish a ranking for species that are routinely released to the environment for a projected nuclear power growth scenario. Unlike comparisons made to existing standards, which are subject to frequent revision, the ranking of releases can be used to form a more logical basis for identifying the areas where further development of control technology could be required. This report describes projections of releases for several fuel cycle scenarios, identifies areas where alternative control technologies may be implemented, and discusses the available alternative control technologies. The release factors were used in a computer code system called ENFORM, which calculates the annual release of any species from any part of the LWR nuclear fuel cycle given a projection of installed nuclear generation capacity. This survey of fuel cycle releases was performed for three reprocessing scenarios (stowaway, reprocessing without recycle of Pu and reprocessing with full recycle of U and Pu) for a 100-year period beginning in 1977. The radioactivity releases were ranked on the basis of a relative ranking factor. The relative ranking factor is based on the 100-year summation of the 50-year population dose commitment from an annual release of radioactive effluents. The nonradioactive releases were ranked on the basis of dilution factor. The twenty highest ranking radioactive releases were identified and each of these was analyzed in terms of the basis for calculating the release and a description of the currently employed control method. Alternative control technology is then discussed, along with the available capital and operating cost figures for alternative control methods.

  9. 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....

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Understanding EUV mask blank surface roughness induced LWR and associated roughness requirement

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Pei-Yang [Intel Corp., Santa Clara, CA (United States); Zhang, Guojing [Intel Corp., Santa Clara, CA (United States); Gullickson, Eric M. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Goldberg, Kenneth A. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Benk, Markus P. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2015-03-01

    Extreme ultraviolet lithography (EUVL) mask multi-layer (ML) blank surface roughness specification historically comes from blank defect inspection tool requirement. Later, new concerns on ML surface roughness induced wafer pattern line width roughness (LWR) arise. In this paper, we have studied wafer level pattern LWR as a function of EUVL mask surface roughness via High-NA Actinic Reticle Review Tool. We found that the blank surface roughness induced LWR at current blank roughness level is in the order of 0.5nm 3σ for NA=0.42 at the best focus. At defocus of ±40nm, the corresponding LWR will be 0.2nm higher. Further reducing EUVL mask blank surface roughness will increase the blank cost with limited benefit in improving the pattern LWR, provided that the intrinsic resist LWR is in the order of 1nm and above.

  15. PIE of nuclear grade SiC/SiC flexural coupons irradiated to 10 dpa at LWR temperature

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    Silicon carbide fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230–340°C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials are chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC)-coated Hi-NicalonTM Type-S (HNS), TyrannoTM SA3 (SA3), and SCS-UltraTM (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young’s modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young’s moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

  16. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-15

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

  17. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-15

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions need to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

  18. 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.

  19. Assessment of LWR piping design loading based on plant operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Svensson, P. O.

    1980-08-01

    The objective of this study has been to: (1) identify current Light Water Reactor (LWR) piping design load parameters, (2) identify significant actual LWR piping loads from plant operating experience, (3) perform a comparison of these two sets of data and determine the significance of any differences, and (4) make an evaluation of the load representation in current LWR piping design practice, in view of plant operating experience with respect to piping behavior and response to loading.

  20. LWR decay heat calculations using a GRS improved ENDF/B-6 based ORIGEN data library

    Energy Technology Data Exchange (ETDEWEB)

    Hesse, U.; Hummelsheim, K.I.; Kilger, R.; Moser, F.E.; Langenbuch, S. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, Garching (Germany)

    2008-07-01

    The known ORNL ORIGEN code is widely spread over the world for inventory, activity and decay heat tasks and is used stand-alone or implemented in activation, shielding or burn-up systems. More than 1000 isotopes with more than six coupled neutron capture and radioactive decay channels are handled simultaneously by the code. The characteristics of the calculated inventories, e.g., masses, activities, neutron and photon source terms or the decay heat during short or long decay time steps are achieved by summing over all isotopes, characterized in the ORIGEN libraries. An extended nuclear GRS-ORIGENX data library is now developed for practical appliance. The library was checked for activation tasks of structure material isotopes and for actinide and fission product burn-up calculations compared with experiments and standard methods. The paper is directed to the LWR decay heat calculation features of the new library and shows the differences of dynamical and time integrated results of Endf/B-6 based and older Endf/B-5 based libraries for decay heat tasks compared to fission burst experiments, ANS curves and some other published data. A multi-group time exponential evaluation is given for the fission burst power of {sup 235}U, {sup 238}U, {sup 239}Pu and {sup 241}Pu, to be used in quick LWR reactor accident decay heat calculation tools. (authors)

  1. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  2. 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

  3. 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.

  4. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  5. 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)

  6. 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.

  7. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

  8. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

  9. Investigation of valve failure problems in LWR power plants

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

  10. Nondestructive evaluation of LWR spent fuel shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Ballard, D.W.

    1978-02-01

    An analysis of nondestructve testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined. A comprehensive bibliography of current NDT methods for cask evaluation in the USA, Great Britain, Japan and West Germany was compiled for this study.

  11. Basic Design of a LWR Fuel Compatibility Test Facility (PLUTO)

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Chun, Se Young; Kim, Bok Deuk; Park, Jong Kuk; Chun, Tae Hyun; Kim, Hyoung Kyu; Oh, Dong Seok

    2009-04-15

    KAERI is performing a project for developing a compatibility test facility and the relevant technology for an LWR fuel assembly. It includes the compatibility test and the long term wear test for dual fuel assemblies, and the pressure drop test, uplift force test, flow-induced vibration test, damping test, and the debris filtering capability test for a single fuel assembly. This compatibility test facility of the fuel assemblies is named PLUTO from Performance Test Facility for Fuel Assembly Hydraulics and Vibrations. The PLUTO will be basically constructed for a PWR fuel assembly, and it will be considered to test for the fuel assemblies of other reactors.

  12. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

    Energy Technology Data Exchange (ETDEWEB)

    Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

  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. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 4: External Pressurizer Surge Line Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  15. APT Blanket System Loss-of-Coolant Accident Based on Initial Conceptual Design - Case 5: External RHR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.

  16. APT Blanket System Loss-of-Coolant Accident (LOCA) Analysis Based on Initial Conceptual Design - Case 3: External HR Break at Pump Outlet without Pump Trip

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal (HR) system. These simulations were performed for the Preliminary Safety Analysis Report.

  17. APT Blanket System Loss-of-Coolant Analysis Based on Initial Conceptual Design - Case 2: External HR Break HR Break at Pump Outlet with Pump Trip

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.

  18. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  19. FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

  20. Research Progress of Decontamination Process and its Corrosion Effect on Primary Coolant Systems of Nuclear Reactor%反应堆一回路系统去污工艺及其对结构材料腐蚀的影响

    Institute of Scientific and Technical Information of China (English)

    谭昭怡; 李烨; 孙宇; 汪小琳; 张东

    2012-01-01

    采用化学去污工艺可降低反应堆一回路冷却系统周围辐射场.总结了近年来反应堆一回路冷却系统去污工艺和去污试剂对结构材料的腐蚀影响的研究成果,并建议后续研究方向.%Radiation field intensity in the primary coolant system of water-cooled reactors could be reduced by chemical decontamination process.Thus,the recent research progresses of the decontamination process and its corrosion effect on structural materials of the primary coolant systems were summarized in this paper.

  1. 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.

  2. 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.

  3. 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.

  4. 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.)

  5. Characteristics Data Base: Programmer's guide to the LWR Quantities Data Base

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.E. (DataPhile, Inc., Knoxville, TN (USA)); Moore, R.S. (Automated Sciences Group, Inc., Oak Ridge, TN (USA))

    1990-08-01

    The LWR Quantities Data Base is a menu-driven PC data base developed as part of OCRWM's waste, technical data base on the characteristics of potential repository wastes, which also includes non-LWR spent fuel, high-level and other materials. This programmer's guide completes the documentation for the LWR Quantities Data Base, the user's guide having been published previously. The PC data base itself may be requested from the Oak Ridge National Laboratory, using the order form provided in Volume 1 of publication DOE/RW-0184.

  6. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  7. 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.

  8. Irradiation effects on thermal properties of LWR hydride fuel

    Science.gov (United States)

    Terrani, Kurt; Balooch, Mehdi; Carpenter, David; Kohse, Gordon; Keiser, Dennis; Meyer, Mitchell; Olander, Donald

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

  9. Development of PIE techniques for irradiated LWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-09-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  10. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    Energy Technology Data Exchange (ETDEWEB)

    Katayama, Y.B.

    1979-04-01

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25/sup 0/C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of /sup 137/Cs and /sup 239 +240/Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures.

  11. 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.

  12. Metholology for the selection of LWR safety R and D projects. Phase I, status report

    Energy Technology Data Exchange (ETDEWEB)

    El-Sheikh, K. A.

    1980-03-01

    The objective of the LWR R and D Selection Methodology Program is to develop and demonstrate an R and D selection methodology appropriate for LWR safety technology. This report documents the development work from the program beginning in April, 1979 to the end of Fiscal Year 1979. The scope of work for this period included three tasks; methodology review (Task 1), measures development (Task 2), and methodology development for the first phase of application (Task 3). The accomplishments of these tasks are presented.

  13. 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.

  14. 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.

  15. 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.

  16. 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...

  17. 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...

  18. 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.)

  19. 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)

  20. 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.

  1. 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.

  2. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-25

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  3. Hybrid reprocessing technology of fluoride volatility and solvent extraction. New reprocessing technology, FLUOREX, for LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Fumio [Hitachi Ltd., Ibaraki (Japan)

    2002-11-01

    Hybrid Process of Fluoride Volatility and Solvent Extraction (FLUOREX) has been objected to develop a low cost reprocessing technology for collection of U and MOX (mixture U and Pu) in LWR fuel cycle. Outline, characteristics, technologies, problems and material balance of FLUOREX are explained. LWR spent fuel consists of about 96% U, 1% Pu and about 3% fission products (FP) and minor actinides (MA). FLUOREX method is hybrid system, which isolates about 90% U at high speed and refines by fluoride volatility process and residue about 10% U, Pu, MA and FP are processed by PUREX method after dissolution in acid. The special features are low cost by small type and lightweight, stable without gas Pu and stop of fluorine gas, reducing load of environment, resistance of nuclear proliferation, application of technologies demonstrated and flexible method for fast reactor. Three problems for development are selective fluoridation of U, transportation of oxides in the fluoride residue and dissolution of transported oxides. The preliminary examination of plan showed 800GWD/t processing volume, 200 day/year operation day, about 51 ten-thousand cubic meter volume of plant, about 1/3 Rokkasho reprocessing plant. (S.Y.)

  4. Uranium nitride as LWR TRISO fuel: Thermodynamic modeling of U-C-N

    Science.gov (United States)

    Besmann, Theodore M.; Shin, Dongwon; Lindemer, Terrence B.

    2012-08-01

    TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will likely need to be UN instead of UO2. In support of the necessary development effort for this new fuel system, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide followed by nitriding, will be in equilibrium with carbon within the TRISO particle, and will react with minor actinides and fission products. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Measurements were used to confirm an ideal solution model of UN and UC adequately represents the UC1-xNx phase. Agreement with the data was significantly improved by effectively adjusting the Gibbs free energy of UN by +12 kJ/mol. This also required adjustment of the value for the sesquinitride by +17 kJ/mol to obtain agreement with phase equilibria. The resultant model together with reported values for other phases in the system was used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.

  5. Robot Kinematics Identification: KUKA LWR4+ Redundant Manipulator Example

    Science.gov (United States)

    Kolyubin, Sergey; Paramonov, Leonid; Shiriaev, Anton

    2015-11-01

    This work is aimed at a comprehensive discussion of algorithms for the kinematic parameters identification of robotic manipulators. We deal with an open-loop geometric calibration task, when a full 6D robot's end-effector pose is measured. Effective solutions of such a task is of high interest in many practical applications, because it can dramatically improve key robot characteristics. On the first step, we select optimal calibration configurations. A comparative analysis of three different algorithms and two observability indexes used for numerical optimization is provided. Afterwards, using the acquired and pre-processed experimental data we identify modified Denavit-Hartenberg parameters of the manipulator. Estimates are obtained resolving original nonlinear forward kinematics relations. Finally, we compare nominal and calibrated geometric parameters and show how much deviations in these parameters affect robot positioning accuracy. To the best of our knowledge, such integrated efforts are new for the KUKA LWR4+ robot and Nikon K610 optical coordinate measuring machine (CMM), which were used in the study. Discussion of practical issues on how to organise the experiment is an additional contribution of this work. The proposed procedure is highly automated and can be implemented to improve manipulator's performance on a periodic basis.

  6. Democratic People`s Republic of Korea LWR project status

    Energy Technology Data Exchange (ETDEWEB)

    Mulligan, J.B.

    1996-07-01

    In October 1994, at Geneva, the United States and the Democratic People`s Republic of Korea (DPRK) signed an Agreed Framework as a first step toward resolving international concerns about nuclear activities in the DPRK. This Agreement, when implemented, will ultimately lead to the complete dismantlement of those aspects of the DPRK`s nuclear program, including reprocessing-related facilities, that have undermined the viability of the international nuclear non-proliferation regime and the stability of the Asia-Pacific region. The essence of the Agreement is that the DPRK will take near-term action to cease the activities of concern and permit some International Atomic Energy Agency (IAEA) verification inspection. In the future, it will dismantle its production reactors and accept full-scope IAWA safeguards. In return, the United Stated agreed to lead an international effort to supply the DPRK with light-water reactors which are less of proliferation concern than are graphite-moderated production reactors. Until the first LWR is in operation the DPRK will receive shipments of heavy oil to replace the energy lost by shutting down the production reactors.

  7. 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

  8. 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.

  9. 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.)

  10. 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.

  11. 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)

  12. Improved LWR Cladding Performance by EPD Surface Modification Technique

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Sridharan, Kumar

    2012-11-26

    This project will utilize the electro-phoretic deposition technique (EPD) in conjunction with nanofluids to deposit oxide coatings on prototypic zirconium alloy cladding surfaces. After demonstrating that this surface modification is reproducible and robust, the team will subject the modified surface to boiling and corrosion tests to characterize the improved nucleate boiling behavior and superior corrosion performance. The scope of work consists of the following three tasks: The first task will employ the EPD surface modification technique to coat the surface of a prototypic set of zirconium alloy cladding tube materials (e.g. Zircaloy and advanced alloys such as M5) with a micron-thick layer of zirconium oxide nanoparticles. The team will characterize the modified surface for uniformity using optical microscopy and scanning-electron microscopy, and for robustness using standard hardness measurements. After zirconium alloy cladding samples have been prepared and characterized using the EPD technique, the team will begin a set of boiling experiments to measure the heat transfer coefficient and critical heat flux (CHF) limit for each prepared sample and its control sample. This work will provide a relative comparison of the heat transfer performance for each alloy and the surface modification technique employed. As the boiling heat transfer experiments begin, the team will also begin corrosion tests for these zirconium alloy samples using a water corrosion test loop that can mimic light water reactor (LWR) operational environments. They will perform extended corrosion tests on the surface-modified zirconium alloy samples and control samples to examine the robustness of the modified surface, as well as the effect on surface oxidation

  13. Feasibility study on the development of advanced LWR fuel technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

    1997-07-01

    Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

  14. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  15. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  16. Effects of Listening While Reading (LWR on Swahili Reading Fluency and Comprehension

    Directory of Open Access Journals (Sweden)

    Filipo Lubua

    2016-10-01

    Full Text Available A number of studies have examined the contribution of technology in teaching such languages as English, French, and Spanish, among many others. Contrarily, most LCTL’s, have received very little attention. This study investigates if listening while reading (LWR may expedite Swahili reading fluency and comprehension. The study employed the iBook Author tool to create weekly mediated and interactive reading texts, with comprehension exercises, which were eventually used to collect descriptive and qualitative data from four Elementary Swahili students. Participants participated in a seven week reading program, which provided them with some kind of directed self-learning, and met with the instructor for at least 30 minutes every week for observation and more reading activities. The teacher recorded their reading scores, and a number of themes on how LWR influenced reading fluency and comprehension are discussed here. It shows that participants have a positive attitude towards LWR and they suggest it for all the reading classes.

  17. 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

  18. 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

  19. 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

  20. In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures

    Science.gov (United States)

    Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.

    2017-02-01

    The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

  1. LWR fuel recycle program quarterly progress report, April--June 1977

    Energy Technology Data Exchange (ETDEWEB)

    Jarrett, J.H. (comp.)

    1977-08-01

    The LWR Fuel Recycle Program is designed to provide information needed by industry to close the back end of the commercial light water reactor (LWR) fuel cycle. Included in this program are activities in support of specific design studies as well as activities with more general application to fuel recycle technology: economic and environmental studies; spent fuel receipt and storage; head-end processes; off-gas treatment; purex process (solvent extraction); finishing processes; waste management; environmental effects; and general support. 11 figures, 7 tables. (DLC)

  2. Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

    Energy Technology Data Exchange (ETDEWEB)

    Miller, W.J.

    1980-05-01

    This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan.

  3. 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.

  4. 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.

  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. Extending dry storage of spent LWR fuel for 100 years.

    Energy Technology Data Exchange (ETDEWEB)

    Einziger, R. E.

    1998-12-16

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and

  7. 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

    To prevent the appearance of the conditions for resonance interaction between the fluid flow and the reactor internals (RI), fuel rod (FR ) and fuel assemblies (FA) it is necessary to de-tune Eigen frequency of coolant pressure oscillations (EFCPO) and natural frequency of mechanical element's oscillations and also of the system which is formed by the comprising of these elements. Other words it is necessary to de-tune acoustic resonance frequency and natural frequencies of RI, FR and FA. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of the coolant outside of which there is no resonant interaction with structure vibrations. The presented work is devoted to finding the solution of this problem. There are results of an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. Abnormal growth of intensity of pressure pulsations in a mode with definite value of reactor capacity have been found out by measurements on VVER - 1000 reactor. This phenomenon has been found out casually and its original reason had not been identified. Paper shows that disappearance of this effect could be reached by realizing outlet of EFCPO from so-called, pass bands of frequencies (PBF). PBF is located symmetrical on both parties from frequency of own oscillations of FA. Methods, algorithms of calculations and quantitative estimations are developed for EFCPO, Q and PBF in various modes of operation NPP with VVER-1000. Results of calculations allow specifying area of resonant interaction EFCPO with vibrations of FR, FA and a basket of reactor core. For practical realization of the received results it is offered to make corresponding additions to the design documentation and maintenance instructions of the equipment of the NPP with VVER-1000. The improvement of these documents

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. Stress analysis for the candidate of lower end fitting of advanced LWR fuel using FEM

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. S.; Moon, Y. C. [Korea University of Technology and Education, Chonan (Korea, Republic of); Kim, H. K. [Korea Nuclear Fuel Company, Taejon (Korea, Republic of)

    2002-10-01

    The geometric modeling has been conducted for the candidate of advanced LWR fuel using the three-dimensional solid modeler. Then the three-dimensional stress analysis using MSC/NASTRAN has been performed. The evaluation for the mechanical integrity of the candidate has been performed based on the stress distribution obtained from the finite elements analysis.

  14. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

  15. Mathematical Description of Wafer-1, a Three-Dimensional Code for LWR Fuel Performance Analysis

    DEFF Research Database (Denmark)

    Kjær-Pedersen, Niels

    1975-01-01

    This article describes in detail the mathematical formulation used in the WAFER-1 code, which is presently used for three-dimensional analysis of LWR fuel pin performance. The code aims at a prediction of the local stress-strain history in the cladding, especially with regard to the ridging pheno...

  16. 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

    conditions, the model was scaled and compared to nominal values of a 4500 MWt ESBWR. The design of the reduced order model controller was mathematically based on the theory of an extended Kalman filter, whose algorithm allows to carry out an advanced control of the system. The estimator uses the equivalent electrical model, which was developed from the system. The estimator uses the equivalent electrical model, which was developed from the analogies between electrical current and voltage with coolant flows and pressure drops. (Author)

  17. 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.

  18. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    Science.gov (United States)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  19. Icare/Cathare coupling: three-dimensional thermal hydraulics of severe LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, V.; Fichot, F. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, DRS, 92 (France); Boudier, P.; Parent, M. [CEA Grenoble, Dir. des Reacteurs Nucleaires, DRN, 38 (France); Roser, R. [Communication et Systemes Systemes d' Information, CS SI, 38 - Fontaine (France)

    2001-07-01

    In the phenomenology of severe LWR accidents considered in safety studies, the accidental sequences can be divided into three phases: the initial phase, where no severe damage of fuel or control rods and structures occurs; the early core degradation phase, where limited material melting and relocation takes place; and the late core degradation phase during which substantial material relocation happens, molten pools and debris beds can form and corium may fall into the lower plenum and, in case of vessel failure, come into the containment. The CATHARE2 code is a system code which has been developed by CEA for IPSN, EDF and FRAMATOME to describe the thermal-hydraulics behavior of a whole PWR circuit during the first of these three phases, with a core degradation model limited to clad rupture. The ICARE2 code, developed by IPSN, allows the complete description of early and late core degradation phases, with a thermal-hydraulics model limited to the vessel, initial and boundary conditions being provided by a system code. The aim of this paper is to present the main features of the new version of the coupling, ICARE/CATHARE V2. First, the general characteristics of ICARE2 V3mod1 and CATHARE2 V1.5 standard codes, dealing with physical models and numerical aspects, are described. Second, the technical features of the coupling between the two codes are detailed. At last, some results of ICARE/CATHARE V2 calculations are presented which demonstrate the ability of the code to simulate a severe accident in a PWR and notably to describe multi-dimensional effects occurring in the core during the LOCA and degradation phases. (authors)

  20. 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)

  1. 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)

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels

    Science.gov (United States)

    Chopra, O. K.; Rao, A. S.

    2011-02-01

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation) of these steels, and degrades their fracture properties. Irradiation-assisted stress corrosion cracking (IASCC) is another degradation process that affects LWR internal components exposed to neutron radiation. The existing data on irradiated austenitic SSs were reviewed to evaluate the effects of key parameters such as material composition, irradiation dose, and water chemistry on IASCC susceptibility and crack growth rates of these materials in LWR environments. The significance of microstructural and microchemistry changes in the material on IASCC susceptibility is also discussed. The results are used to determine (a) the threshold fluence for IASCC and (b) the disposition curves for cyclic and IASCC growth rates for irradiated SSs in LWR environments.

  8. 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)

  9. 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.

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

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

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

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  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. Review of Severe Accident Phenomena in LWR and Related Severe Accident Analysis Codes

    Directory of Open Access Journals (Sweden)

    Muhammad Hashim

    2013-04-01

    Full Text Available Firstly, importance of severe accident provision is highlighted in view of Fukushima Daiichi accident. Then, extensive review of the past researches on severe accident phenomena in LWR is presented within this study. Various complexes, physicochemical and radiological phenomena take place during various stages of the severe accidents of Light Water Reactor (LWR plants. The review deals with progression of the severe accidents phenomena by dividing into core degradation phenomena in reactor vessel and post core melt phenomena in the containment. The development of various computer codes to analyze these severe accidents phenomena is also summarized in the review. Lastly, the need of international activity is stressed to assemble various severe accidents related knowledge systematically from research organs and compile them on the open knowledge base via the internet to be available worldwide.

  19. FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

    Directory of Open Access Journals (Sweden)

    WEON-JU KIM

    2013-08-01

    Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

  20. 反应堆一回路系统优化设计方案的可行性验证%Feasibility Test for Reactor Coolant System Optimized Design Scheme

    Institute of Scientific and Technical Information of China (English)

    陈磊; 阎昌琪; 王建军

    2014-01-01

    采用优选运行参数和结构参数的方法,可达到降低核动力装置尺寸的目的。在优化设计方案投入制造前,有必要研究其在设计基准事故下的响应特性,以检验优化方案的可行性。采用 REL A P5/M OD3.2程序研究现有一回路系统优化方案在完全失去厂外电、主给水丧失和小破口失水事故下的响应特性,并将安全设计准则参数与母型对比。结果表明:针对所研究的3种设计基准事故,优化方案各主要安全准则参数满足设计要求;优化方案可成功抵御这3类设计基准事故。%The size of a nuclear component could be reduced by optimum selections of the operational and structural parameters .Before an optimized design scheme is manu‐factured ,it is necessary to obtain its transient behaviors and verify its feasibility under design basis accidents .In this work ,the RELAP5/MOD3.2 code was employed to simulate the transient characteristics of a proposed optimized scheme under the complete loss of off‐site power ,loss of feedwater and small break loss of coolant accidents ,and the safety criteria were compared with the prototype reactor design . The simulation results indicate that the safety criteria of the optimized scheme satisfy the design requirements ,and the safety of the optimized scheme can be guaranteed in those three accidents .

  1. Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-08-29

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U3Si2 at LWR conditions. The fission gas behavior of U3Si2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U3Si2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U3Si2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U3Si2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

  2. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  3. Draft report: a selection methodology for LWR safety R and D programs and proposals

    Energy Technology Data Exchange (ETDEWEB)

    Husseiny, A. A.; Ritzman, R. L.

    1980-03-01

    The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application.

  4. Comment: collection of assay data on isotopic composition in LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Naito, Yoshitaka; Kurosawa, Masayoshi; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Many assay data of LWR spent fuels have been collected from reactors in the world and some of them are already stored in the database SFCOMPO which was constructed on a personal computer IBM PC/AT. On the other hand, Group constant libraries for burnup calculation code ORIGEN-II were generated from the nuclear data file JENDL3.2. These libraries were evaluated by using the assay data in SFCOMPO. (author)

  5. Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-08-29

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions need to be well-understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U3Si2 at LWR conditions. The fission gas behavior of U3Si2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U3Si2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U3Si2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U3Si2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

  6. Results of the LIRES Round Robin test on high temperature reference electrodes for LWR applications

    Energy Technology Data Exchange (ETDEWEB)

    Bosch, R.W. [SCK.CEN, Nuclear Research Centre Belgium, Boeretang 200, B-2400 Mol (Belgium); Nagy, G. [Magyar Tudomanyos Akademia KFKI Atomenergia Kutatointezet, AEKI, Konkoly Thege ut 29-33, 1121 Budapest (Hungary); Feron, D. [CEA Saclay, 91191 Gif-Sur-Yvette Cedex (France); Navas, M. [CIEMAT, Edificio 30, Dpto. Fision Nuclear, Avda. Complutense 22, 28040 Madrid, (Spain); Bogaerts, W. [KU Leuven, Kasteelpark Arenberg 31, B-3001 Leuven (Belgium); Karnik, D. [Nuclear Research Institute, NRI, Rez (Czech Republic); Dorsch, T. [Framatone ANP, Inc., Charlotte, North Carolina (United States); Molander, A. [Studsvik AB SE-611 82 Nykoeping (Sweden); Maekelae, K. [Materials and Structural Integrity, VTT Technical Research Centre of Finland, Kemistintie 3, P.O. Box 1704, FIN-02044 VTT (Finland)

    2004-07-01

    A European sponsored research project has been started on 1 October 2000 to develop high temperature reference electrodes that can be used for in-core electrochemical measurements in Light Water Reactors (LWR's). This LIRES-project (Development of Light Water Reactor Reference Electrodes) consists of 9 partners (SCK-CEN, AEKI, CEA, CIEMAT, KU Leuven, NRI Rez, Framatone ANP, Studsvik Nuclear and VTT) and will last for four years. The main objective of this LIRES project is to develop a reference electrode, which is robust enough to be used inside a LWR. Emphasize is put on the radiation hardness of both the mechanical design of the electrode as the proper functioning of the electrode. A four steps development trajectory is foreseen: (1) To set a testing standard for a Round Robin, (2) To develop different reference electrodes, (3) To perform a Round Robin test of these reference electrodes followed by selection of the best reference electrode(s), (4) To perform irradiation tests under appropriate LWR conditions in a Material Test Reactor (MTR). Four different high temperature reference electrodes have been developed and are being tested in a Round Robin test. These electrodes are: A Ceramic Membrane Electrode (CME), a Rhodium electrode, an external Ag/AgCl electrode and a Palladium electrode. The presentation will focus on the results obtained with the Round Robin test. (authors)

  7. 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.

  8. 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.

  9. 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...

  10. 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...

  11. 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…

  12. 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.

  13. 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...

  14. 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...

  15. 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

  16. Data center cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J; Dang, Hien P; Parida, Pritish R; Schultz, Mark D; Sharma, Arun

    2015-03-17

    A data center cooling system may include heat transfer equipment to cool a liquid coolant without vapor compression refrigeration, and the liquid coolant is used on a liquid cooled information technology equipment rack housed in the data center. The system may also include a controller-apparatus to regulate the liquid coolant flow to the liquid cooled information technology equipment rack through a range of liquid coolant flow values based upon information technology equipment temperature thresholds.

  17. 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.

  18. LWR design decision methodology: Phase II. Final report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-01-01

    Techniques were identified to augment existing design process at the component and system level in order to optimize cost and safety between alternative system designs. The method was demonstrated using the Surry Low Pressure Injection System (LPIS). Three possible backfit options were analyzed for the Surry LPIS, assessing the safety level of each option and estimating the acquisition and installation costs for each. (DLC)

  19. 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)

  20. 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.

  1. Opportunities for the LWR ATF materials development program to contribute to the LBE-cooled ADS materials qualification program

    Science.gov (United States)

    Gong, Xing; Li, Rui; Sun, Maozhou; Ren, Qisen; Liu, Tong; Short, Michael P.

    2016-12-01

    Accelerator-driven systems (ADS) are a promising approach for nuclear waste disposal. Nevertheless, the principal candidate materials proposed for ADS construction, such as the ferritic/martensitic steel, T91, and austenitic stainless steels, 316L and 15-15Ti, are not fully compatible with the liquid lead-bismuth eutectic (LBE) coolant. Under some operating conditions, liquid metal embrittlement (LME) or liquid metal corrosion (LMC) may occur in these steels when exposed to LBE. These environmentally-induced material degradation effects pose a threat to ADS reactor safety, as failure of the materials could initiate a severe accident, in which fission products are released into the coolant. Meanwhile, parallel efforts to develop accident-tolerant fuels (ATF) in light water reactors (LWRs) could provide both general materials design philosophies and specific material solutions to the ADS program. In this paper, the potential contributions of the ATF materials development program to the ADS materials qualification program are evaluated and discussed in terms of service conditions and materials performance requirements. Several specific areas where coordinated development may benefit both programs, including composite materials and selected coatings, are discussed.

  2. 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.

  3. PANDA: A Multipurpose Integral Test Facility for LWR Safety Investigations

    Directory of Open Access Journals (Sweden)

    Domenico Paladino

    2012-01-01

    Full Text Available The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the support of a large number of organizations (regulatory bodies, technical dupport organizations, national laboratories, electric utilities, industries worldwide. The paper provides an overview of the research programs performed in the PANDA facility in relation to BWR containment systems and those planned for PWR containment systems.

  4. 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.

  5. 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).

  6. 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

  7. Optimization of accelerator-driven technology for LWR waste transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, C.D.

    1996-12-31

    The role of accelerator-driven transmutation technology is examined in the context of the destruction of actinide waste from commercial light water reactors. It is pointed out that the commercial plutonium is much easier to use for entry-level nuclear weapons than weapons plutonium. Since commercial plutonium is easier to use, since there is very much more of it already, and since it is growing rapidly, the permanent disposition of commercial plutonium is an issue of greater importance than weapons plutonium. The minor actinides inventory, which may be influenced by transmutation, is compared in terms of nuclear properties with commercial and weapons plutonium and for possible utility as weapons material. Fast and thermal spectrum systems are compared as means for destruction of plutonium and the minor actinides. it is shown that the equilibrium fast spectrum actinide inventory is about 100 times larger than for thermal spectrum systems, and that there is about 100 times more weapons-usable material in the fast spectrum system inventory compared to the thermal spectrum system. Finally it is shown that the accelerator size for transmutation can be substantially reduced by design which uses the accelerator-produced neutrons only to initiate the unsustained fission chains characteristic of the subcritical system. The analysis argues for devoting primary attention to the development of thermal spectrum transmutation technology. A thermal spectrum transmuter operating at a fission power of 750-MWth fission power, which is sufficient to destroy the actinide waste from one 3,000-MWth light water reactor, may be driven by a proton beam of 1 GeV energy and a current of 7 mA. This accelerator is within the range of realizable cyclotron technology and is also near the size contemplated for the next generation spallation neutron source under consideration by the US, Europe, and Japan.

  8. Station Blackout Initiated Event Chronology in LWR/HWR NPP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Since the crisis at Fukushima nuclear power plants, a severe accident progression has been recognized as a very important area for an accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of a severe accident progression among the typical pressurized water reactor (PWR), boiling water reactor (BWR) and pressurized heavy water reactor (PHWR). The OPR 1000-like (ABB-CE type PWR), Peach Bottom-like (BWR/4 RCS with a MARK I Containment), and Wolsong1-like (CANDU6 type) plants are selected as reference plants of typical 1000 MWe PWR, 1140MWe BWR, and 600 MWe PHWR, respectively. The design parameters of these plants are quite different. Some of the major different design features of CANDU6 plant from other light water reactors, in terms of a severe accident, are that the plant adopts a duel primary heat transport system and has an additional amount of cooling water in the calandria vessel (calandria tank, CT) and calandria vault (CV). Another feature is that the CT is always submerged in water because the CV is flooded during normal operation. The containment (reactor building, R/B) failure pressure of the CANDU6 plant is considerably lower than that of the typical PWR or BWR4/MARK-I. The containment vessel free volume of MARK-I is much smaller than that of the PWR or CANDU6 plant. Since there is no steam generator (SG) or passive cooling system, the amount of cooling water inventory in BWR4 is relatively less than other plants. Meanwhile the minimum available time of battery power against station blackout (SBO) accident is different among plant types: six hours for BWR4 and four hours for 1000MWe PWR. Therefore, plant responses against the severe core damage scenarios like Fukushima accident are expected to be much different. By identifying plant response signatures, the appropriate correction actions can be developed as part of severe accident management. A SBO scenario, where all off-site power is lost

  9. PANDA: A Multipurpose Integral Test Facility for LWR Safety Investigations

    OpenAIRE

    2012-01-01

    The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the supp...

  10. MCOR - Monte Carlo depletion code for reference LWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

    2011-04-15

    Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally

  11. Technical program to study the benefits of nonlinear analysis methods in LWR component designs. Technical report TR-3723-1

    Energy Technology Data Exchange (ETDEWEB)

    Raju, P. P.

    1980-05-01

    This report summarizes the results of the study program to assess the benefits of nonlinear analysis methods in Light Water Reactor (LWR) component designs. The current study reveals that despite its increased cost and other complexities, nonlinear analysis is a practical and valuable tool for the design of LWR components, especially under ASME Level D service conditions (faulted conditions) and it will greatly assist in the evaluation of ductile fracture potential of pressure boundary components. Since the nonlinear behavior is generally a local phenomenon, the design of complex components can be accomplished through substructuring isolated localized regions and evaluating them in detail using nonlinear analysis methods.

  12. 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.

  13. DESIK final report. Energy wise design and regulation of the secondary side of indirect cooling systems with natural coolants; DESIK slutrapportering. Energirigtig design og regulering af sekundaersiden pae indirekte koeleanlaeg med naturlige koelemidler

    Energy Technology Data Exchange (ETDEWEB)

    Jakobsen, Arne

    2006-02-15

    The project's aim was to produce knowledge and tools to facilitate the process of implementing energy efficient secondary cooling systems, or merely to avoid overconsumption of energy on account of insufficient relevant professional background. The project has been communicated as a PC tool, which can be ordered from aj(commercial at)ipu.dk. Project focus has been on some general aspects of secondary systems as well as two scopes of application: supermarket cooling systems and air conditioning of office buildings. (BA)

  14. On the Diffusion Coefficient of Two-step Method for LWR analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Deokjung; Choi, Sooyoung [UNIST, Ulsan (Korea, Republic of); Smith, Kord S. [Massachusetts Institute of Technology, Cambridge (United States)

    2015-10-15

    The few-group constants including diffusion coefficients are generated from the assembly calculation results. Once the assembly calculation is done, the cross sections (XSs) are spatially homogenized, and a critical spectrum calculation is performed in order to take into account the neutron leakages of the lattice. The diffusion coefficient is also generated through the critical spectrum calculation. Three different methods of the critical spectrum calculation such as B1 method, P1 method, and fundamental mode (FM) calculation method are considered in this paper. The diffusion coefficients can also be affected by transport approximations for the transport XS calculation which is used in the assembly transport lattice calculation in order to account for the anisotropic scattering effects. The outflow transport approximation and the inflow transport approximation are investigated in this paper. The accuracy of the few group data especially the diffusion coefficients has been studied to optimize the combination of the transport correction methods and the critical spectrum calculation methods using the UNIST lattice physics code STREAM. The combination of the inflow transport approximation and the FM method is shown to provide the highest accuracy in the LWR core calculations. The methodologies to calculate the diffusion coefficients have been reviewed, and the performances of them have been investigated with a LWR core problem. The combination of the inflow transport approximation and the fundamental mode critical spectrum calculation shows the smallest errors in terms of assembly power distribution.

  15. 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.

  16. TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.; Bess, John D.; Housley, Gregory K.

    2016-09-01

    Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment while safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional

  17. 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...

  18. 高温气冷堆数字化规程系统的总体设计%Study on General Design of the Digital Procedure System of High Temperature Gas Coolant Reactor

    Institute of Scientific and Technical Information of China (English)

    冯静阁; 黄晓津

    2015-01-01

    This thesis analyzed digital procedures of the domestic and international nuclear power plant, combined with the own characteristics of the high temperature reactor demonstra-tion project (HTR-PM). The general design of digital procedure system (DPS) of the demonstra-tion project was studied by comprehensively utilizing the multi-disciplinary knowledge including the digital control system (DCS) technology, human factors engineering theory, and computer technology. It provides basis for further study of the system.%文章分析了国内外核电站数字化规程研究现状,结合高温气冷堆示范工程(H T R-P M)的自身特点,综合运用数字化仪控系统(D C S)技术、人因工程理论、计算机技术等多学科知识,对示范工程数字化规程系统的总体设计进行了研究,为该系统的进一步研发提供了基础.

  19. 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

  20. 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.

  1. 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.

  2. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

  3. 二代改进型核电厂严重事故下一回路卸压时机敏感性研究%Sensitivity Analysis on Time of Reactor Coolant System Depressurization under Severe Accident for Generation II+ Nuclear Power Plants

    Institute of Scientific and Technical Information of China (English)

    种毅敏; 杨志义; 石雪垚; 张佳佳; 李春; 倪曼; 徐雨婷

    2015-01-01

    Reactor coolant system (RCS)depressurization is necessary measure for nuclear power plant to mitigate the severe accident,as well as a significant part of the severe accident management guidelines (SAMG).Difference may exist on the time of RCS depressurization in different NPPs.In this paper,based on MAAP4,the sensitivity analysis of time to implement RCS depressurization is performed.A typical integrated computer program,and different effects on mitigation of severe accident are compared.The simulation scenario is typical drill situation of generation II + NPPs,and conclusions can be reference for similar NPPs to implement severe accident management strategy.%一回路卸压是核电厂缓解严重事故的必要手段,也是严重事故管理导则(SAMG)的重要内容,国内核电厂严重事故管理中对一回路卸压的要求并不相同,本文基于典型二代改进型核电厂 SAMG 演练的场景,使用一体化计算程序 MAAP4,对一回路卸压时机进行敏感性分析,比较不同卸压时机对缓解严重事故效果的影响,所给出的结论可为相同类型核电厂制定严重事故管理策略时提供参考。

  4. 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.

  5. 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.

  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. NASAP: a computer code for the evaluation of the Non-proliferation Alternative Systems Assessment Program concepts. Final report in support of Task 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Maul, B.A.

    1979-09-01

    The Non-Proliferation Alternative Systems Assessment Program (NASAP) computer code was developed to calculate the LWR and NASAP choice reactor cost through an arbitrary year T/sub N/. The final cost is arrived at by calculation of cost contributory factors for both LWR and NASAP choice reactors.

  10. Integrated Computational Modeling of Water Side Corrosion in Zirconium Metal Clad Under Nominal LWR Operating Conditions

    Science.gov (United States)

    Aryanfar, Asghar; Thomas, John; Van der Ven, Anton; Xu, Donghua; Youssef, Mostafa; Yang, Jing; Yildiz, Bilge; Marian, Jaime

    2016-10-01

    A mesoscopic chemical reaction kinetics model to predict the formation of zirconium oxide and hydride accumulation light-water reactor (LWR) fuel clad is presented. The model is designed to include thermodynamic information from ab initio electronic structure methods as well as parametric information in terms of diffusion coefficients, thermal conductivities and reaction constants. In contrast to approaches where the experimentally observed time exponents are captured by the models by design, our approach is designed to be predictive and to provide an improved understanding of the corrosion process. We calculate the time evolution of the oxide/metal interface and evaluate the order of the chemical reactions that are conducive to a t 1/3 dependence. We also show calculations of hydrogen cluster accumulation as a function of temperature and depth using spatially dependent cluster dynamics. Strategies to further cohesively integrate the different elements of the model are provided.

  11. Thermochemical prediction of chemical form distributions of fission products in LWR mixed oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1998-06-01

    Radial distribution of chemical forms of fission products (FPs) in LWR mixed oxide (MOX) fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The amounts of fission products generated in the fuel were calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. The effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that the radial distribution of chemical forms depends strongly on the LHR and the O/M ratio, and is not proportional to that of burnup. (author)

  12. Integrated Computational Modeling of Water Side Corrosion in Zirconium Metal Clad Under Nominal LWR Operating Conditions

    Science.gov (United States)

    Aryanfar, Asghar; Thomas, John; Van der Ven, Anton; Xu, Donghua; Youssef, Mostafa; Yang, Jing; Yildiz, Bilge; Marian, Jaime

    2016-11-01

    A mesoscopic chemical reaction kinetics model to predict the formation of zirconium oxide and hydride accumulation light-water reactor (LWR) fuel clad is presented. The model is designed to include thermodynamic information from ab initio electronic structure methods as well as parametric information in terms of diffusion coefficients, thermal conductivities and reaction constants. In contrast to approaches where the experimentally observed time exponents are captured by the models by design, our approach is designed to be predictive and to provide an improved understanding of the corrosion process. We calculate the time evolution of the oxide/metal interface and evaluate the order of the chemical reactions that are conducive to a t 1/3 dependence. We also show calculations of hydrogen cluster accumulation as a function of temperature and depth using spatially dependent cluster dynamics. Strategies to further cohesively integrate the different elements of the model are provided.

  13. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  14. Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, S. J. [ed.

    1980-04-01

    The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation.

  15. 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.

  16. North Korean nuclear issues and the LWR project; analysis of the key technologies

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Koo; Kwack, E. H.; Shin, J. S.; An, J. S.; Lee, J. U.; Kim, H. T.; Kim, J. S.; Yoon, Y. C

    2000-11-01

    Year 2000 will be remembered as an epoch making period between two Koreas. Korean nuclear industries became activated with the KEDO LWR main contracts entering into force in February, 2000. Respective design, manufacturing and construction activities are mobilized in accordance with the total project schedule of about 100 months. What started out as the nuclear power plant design standardization project in the early '80s, is now being implemented as repeat construction of KSNPs at Yonggwang and Ulchin sites as well as at Kumho site in the DPRK. However, the KEDO construction schedule and the past nuclear inconsistency issues are closely linked due to the nonproliferation concerns. In practice, the IAEA must come to the conclusion that the correctness and completeness must be fulfilled before delivery of the first key component for the KEDO LWR unit 1. While the IAEA verification process tends to focus on the nuclear materials accountancy control, longer term objective between two Koreas is bound to take the form of confidence building. It is necessary to analyse the nuclear research and production facilities in order to make proper evaluation of a nation's nuclear capabilities. Close assessment on development status of graphite moderated reactors and their operation history, spent fuel reprocessing facilities, and HEU production capabilities would be essential. In addition, illicit trafficking possibilities should be addressed. Chapter 1 describes the graphite moderated reactors in general; Chapter 2 describes various reprocessing processes and their detection capabilities; Chapter 3 contains possible uranium enrichment processes with their detection capabilities, and Chapter 4 summarizes the international treaties in illicit trafficking control with the IAEA database.

  17. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    Science.gov (United States)

    Schanz, G.; Hagen, S.; Hofmann, P.; Schumacher, G.; Sepold, L.

    1992-06-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400°C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4C), which are leading to extensive low-temperature melt formation around 1200°C. Interrelations between those basic phenomena, resulting for example in cladding deformation ("flowering") and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ("quenching") are determining the evolution paths of fuel element destruction, which are to be identified. A further important task is the abstraction from mechanistic and microstructural details in order to get a rough classification of damage regimes (temperature and extent), a practicable analytical treatment of the materials behaviour, and a basis for decisions in accident mitigation and management procedures.

  18. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  19. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.; Parks, C. V.

    2000-12-11

    This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental

  20. 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.

  1. 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.

  2. 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.)

  3. 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.

  4. 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.

  5. Cooling system with automated seasonal freeze protection

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, Levi A.; Chu, Richard C.; David, Milnes P.; Ellsworth, Jr., Michael J.; Iyengar, Madhusudan K.; Simons, Robert E.; Singh, Prabjit; Zhang, Jing

    2016-05-24

    An automated multi-fluid cooling system and method are provided for cooling an electronic component(s). The cooling system includes a coolant loop, a coolant tank, multiple valves, and a controller. The coolant loop is at least partially exposed to outdoor ambient air temperature(s) during normal operation, and the coolant tank includes first and second reservoirs containing first and second fluids, respectively. The first fluid freezes at a lower temperature than the second, the second fluid has superior cooling properties compared with the first, and the two fluids are soluble. The multiple valves are controllable to selectively couple the first or second fluid into the coolant in the coolant loop, wherein the coolant includes at least the second fluid. The controller automatically controls the valves to vary first fluid concentration level in the coolant loop based on historical, current, or anticipated outdoor air ambient temperature(s) for a time of year.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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)

  13. Vapor pressures of mixtures of CFC-114 with the potential replacement coolants C{sub 4}F{sub 10} and c-C{sub 4}F{sub 8}

    Energy Technology Data Exchange (ETDEWEB)

    Trowbridge, L.D. [Oak Ridge K-25 Site, TN (United States); Otey, M.G. [Paducah Gaseous Diffusion Plant, KY (United States)

    1994-09-01

    The U.S. Enrichment Corporation`s production of isotopically enriched uranium depends solely on two plants which utilize the gaseous diffusion process. This process uses large quantities of CFC-114 as an evaporative coolant. CFC-114, however, will be phased out of production at the end of 1995 due to its potential to deplete stratospheric ozone. A search has been underway for substitutes for a number of years. The initial search (1988-89) for an ozone-friendly, commercially available, chemically compatible substitute yielded two candidates, FC-c318 (c-C{sub 4}F{sub 8}) and FC-3110 (C{sub 4}F{sub 10}). The intended mode of replacing coolant was to stage the new coolant into independent subsystems of the plants, so that some systems would continue to operate on CFC-114, and an increasing number would operate on the new coolant. During that changeover process, the possibility of coolant mixing arises in variety of scenarios. This work was intended to generate sufficient experimental information to be able to predict the vapor pressure of coolant mixtures over the range of operating conditions likely to be found in the diffusion plants. Specifically, vapor pressures were measured over the temperature range 322 to 355 K (120{degrees}F to 180{degrees}F) and over the full range of mole fractions for binary mixtures of CFC-114 with FC-3110, and of CFC-114 with FC-c318.

  14. 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

  15. Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lippek, H.E.; Schuller, C.R.

    1979-03-01

    A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.

  16. 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.

  17. Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States); Mei, Zhigang [Argonne National Lab. (ANL), Argonne, IL (United States); Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-10

    Uranium silicide (U3Si2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U3Si2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U3Si2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuel material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U3Si2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U3Si2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U3Si2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U3Si2 fuel as an accident

  18. Study on the establishment of effective nuclear export system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Koo; So, Dong Sup; Baik, Dae Hyun; Kwack, Eun Ho; Shin, Jang Soo; Yoon, Wan Ki; Park, Wan Soo; Kim, Hyun Tae

    1997-02-01

    To improve Korean nuclear export control system, the modification of the present export license procedure for the nuclear equipment and materials and the classification of control items and their related technologies are required. And it is also necessary to make a database of the original countries who have the right of prior consent. For the efficient export control of LWR items to DPRK, it is desirable to manage the export license scheme of nuclear reactor facility as a total package, and to prepare a control regime for the retransfer of nuclear reactor component such as reactor coolant pump and nuclear fuel whose technologies are not self-reliant. It is especially essential to prepare a systematic procedure for the supply of nuclear equipment and materials to DPRK in order to meet international guidelines of NSG and others through an accord on the nuclear cooperation between Republic of Korea (ROK) and DPRK. The principal elements to be included in the accord are the range of cooperation, the restriction within the peaceful uses, prior consent right in case of retransfer of important nuclear reactor components and of storage, transfer and changes of nuclear fuels, application of safeguards to the supplied Trigger list items, physical protection of nuclear material, requirement of the return of nuclear equipment and materials, and restriction right for the suspension or termination of the agreement. (author). 40 refs., 5 tabs., 8 figs.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. The impact of integrated water management on the Space Station propulsion system

    Science.gov (United States)

    Schmidt, George R.

    1987-01-01

    The water usage of elements in the Space Station integrated water system (IWS) is discussed, and the parameters affecting the overall water balance and the water-electrolysis propulsion-system requirements are considered. With nominal IWS operating characteristics, extra logistic water resupply (LWR) is found to be unnecessary in the satisfaction of the nominal propulsion requirements. With the consideration of all possible operating characteristics, LWR will not be required in 65.5 percent of the cases, and for 17.9 percent of the cases LWR can be eliminated by controlling the stay time of theShuttle Orbiter orbiter.

  5. 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)

  6. Preliminary analysis of modeling of Pars and steam injectors to support long-term operation of LWR passive ECCS using a best estimate thermal-hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J. B.; Sanchez J, J. [UNAM, Facultad de Ingenieria, Circuito Interior s/n, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: jaimebmoraless@gmail.co [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    In order to control the amount of hydrogen produced, during the chemical reaction of steam and strongly overheated zirconium in a LWR severe accident, several control methos have been proposed. In a considerable number of nuclear power plants the initial preferred solution was either the use of an inert containment atmosphere or the use of igniters. However, the use of Pars (Passive autocatalytic recombiners) has been considered an important addition to this set of H{sub 2} controls. Pars have appealing features such as, low maintenance, do not require active components to start operation, and can initiate the H{sub 2} and O{sub 2} recombination at low concentrations. However, under certain conditions these Pars may trigger H{sub 2} ignition leading to deflagration s with possible containment damages. Combustions, poisons and liquid may inhibit or reduce the Pars operation. Advanced LWR designs (generation III + and IV) are also considering the use of Pars as a supporting system to the long term operation of their passive ECCS (Emergency Core Cooling Systems) because they may transport the energy of containment H{sub 2} and O{sub 2} gases to external sinks. Even when the ECCS of modern nuclear power plants are designed to maintain containment conditions under established limits, there are small amounts of O{sub 2} and H{sub 2} produced mainly by H{sub 2}O radiolysis, and after several days of containment isolation, these gases may accumulate in places where the possibility of a H{sub 2} deflagration increases. Therefore, the use of Pars looks initially as an interesting solution to long-term H{sub 2} control during a Dba. However, the H{sub 2} and O{sub 2} recombination is an exothermic process resulting in additional steam and energy releases to the containment. In order to avoid a direct deposition of the steam and energy (Pars produced) into the containment atmosphere, we propose to use them to operate steam injectors that can bring cold water to replenish ECCS

  7. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  8. 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.

  9. 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.

  10. Analysis of fission gas release in LWR fuel using the BISON code

    Energy Technology Data Exchange (ETDEWEB)

    G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

  11. Proceedings of the IAEA specialists` meeting on cracking in LWR RPV head penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Pugh, C.E.; Raney, S.J. [comps.] [Oak Ridge National Lab., TN (United States)

    1996-07-01

    This report contains 17 papers that were presented in four sessions at the IAEA Specialists` meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately.

  12. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    Energy Technology Data Exchange (ETDEWEB)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-10-16

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle.

  13. Fission Product Release and Survivability of UN-Kernel LWR TRISO Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M [ORNL; Ferber, Mattison K [ORNL; Lin, Hua-Tay [ORNL

    2014-01-01

    A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from range calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 m diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated with a TRISO particle as a function of fluence. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by measuring the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers as a function of fluence. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

  14. LWR nuclear fuel bundle data for use in fuel bundle handling

    Energy Technology Data Exchange (ETDEWEB)

    Weihermiller, W.B.; Allison, G.S.

    1979-09-01

    Although increasing numbers of spent light water reactor (LWR) fuel bundles are moved into storage, no handling equipment is set up to manipulate all of the various types of fuel bundles. This report summarizes fuel bundle information of interest to the designer of such handling equipment. Dimensional descriptions are included with discussions of assembly procedure and manufacturer provisions for handling equipment. No attempt is made to make a complete compilation of dimensional information; the number of fuel bundle designs and design revisions makes it impractical. Because the fuel bundle designs are so varied, any equipment intended for handling all types of bundles will have to be designed with flexibility in mind. Besides the ability to manipulate fuel bundles in space, handling equipment may be required to locate an external surface or to position a cutting operation to avoid breaking a fuel rod pressure boundary. Even with the most sophisticated and flexible handling equipment, some situations will require use of the manufacturers' as-built descriptions of individual fuel bundles.

  15. Storage of LWR (light-water-reactor) spent fuel in air

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, L.E.; Charlot, L.A.; Coleman, J.E. (Pacific Northwest Lab., Richland, WA (USA)); Knoll, R.W. (Johnson Controls, Inc., Madison, WI (USA))

    1989-12-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about 34 MWd/kgM), reactor type (pressurized and boiling water reactors), moisture level in the air, and the presence of a high gamma field. In continuing tests with declad spent fuel and nonirradiated UO{sub 2} specimens, oxidation rates were monitored by weight-gain measurements and the microstructures of subsamples taken during the weighing intervals were characterized by several analytical methods. The oxidation behavior indicated by weight gain and time to form powder will be reported in Volume III of this series. The characterization results obtained from x-ray diffractometry, transmission electron microscopy, scanning electron microscopy, and Auger electron spectrometry of oxidized fuel samples are presented in this report. 28 refs., 21 figs., 3 tabs.

  16. 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.

  17. 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.

  18. Study of the reliability of the Auxiliary Feedwater System of a LWR nuclear power plant through the Fault Tree and Bayesian Network; Estudo de confiabilidade do Sistema Auxiliar de Agua de Alimentacao de uma central nuclear a agua leve por arvore de falhas e rede Bayesiana

    Energy Technology Data Exchange (ETDEWEB)

    Lava, Deise Diana

    2016-10-01

    This paper aims to present a study of the reliability of the Auxiliary Feedwater System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10{sup -3}. (author)

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

  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. 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)

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DelCul, Guillermo Daniel [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL

    2009-02-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

  11. Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

    Energy Technology Data Exchange (ETDEWEB)

    Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

    1995-12-31

    Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

  12. Impacts of transient heat transfer modeling on prediction of advanced cladding fracture during LWR LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2016-03-15

    Highlights: • Use of constant heat transfer coefficient for fracture analysis is not sound. • On-time heat transfer coefficient should be used for thermal fracture prediction. • ∼90% of the actual fracture stresses were predicted with the on-time transient h. • Thermal-hydraulic codes can be used to better predict brittle cladding fracture. • Effects of surface oxides on thermal shock fracture should be accounted by h. - Abstract: This study presents the importance of coherency in modeling thermal-hydraulics and mechanical behavior of a solid for an advanced prediction of cladding thermal shock fracture. In water quenching, a solid experiences dynamic heat transfer rate evolutions with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates has been overlooked in the analysis of thermal shock fracture. In this study, we are presenting quantitative evidence against the prevailing use of a constant heat transfer coefficient for thermal shock fracture analysis in water. We conclude that no single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials. The presented results show a remarkable stress prediction improvement up to 80–90% of the actual stress with the use of the surface temperature dependent heat transfer coefficient. For thermal shock fracture analysis of brittle fuel cladding such as oxidized zirconium-based alloy or silicon carbide during LWR reflood, transient subchannel heat transfer coefficients obtained from a thermal-hydraulics code should be used as input for stress analysis. Such efforts will lead to a fundamental improvement in thermal shock fracture predictability over the current experimental empiricism for cladding fracture analysis during reflood.

  13. 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.

  14. Status of the MeLoDIE experiment, an advanced device for the study of the irradiation creep of LWR cladding with full online capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Guimbal, P. [CEA, DEN, DER, Jules Horowitz Reactor, F-13108 St-Paul-lez-Durance (France); Huotilainen, S.; Taehtinen, S. [VTT Technical Research Centre of Finland, Materials performance division, Espoo, FI-02044 VTT (Finland); Thellier, G. [CEA, DEN, DRSN, OSIRIS Reactor, F-91191 Gif-sur-Yvette (France); Villard, J.F. [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, F-13108 St- Paul-lez-Durance (France)

    2015-07-01

    As a prototype of future instrumented material experiments in the Jules Horowitz Reactor (JHR), the MELODIE project was launched in 2009 by the CEA in collaboration with VTT. Being designed as a biaxial creep experiment with online capability, MELODIE is able to apply an online-controlled biaxial loading on a LWR clad sample up to 120 MPa and to perform an online measurement of its biaxial deformation. An important experimental challenge was to perform reliably accurate measurements under the high nuclear heat load of in-core locations while keeping within their tight space. For that purpose, specific sensors were co-designed with and built by IFE Halden. Manufacturing of the MELODIE components was completed one year ago. The complexity of its in-pile section and of the pressurization system requested a step-by-step tuning of the setup. The toughest part of this process dealt with the Diameter gauge which required a partial redesign to take into account unexpected and unwanted electromagnetic interactions with the hosting device. Final cold performance tests of the on-board instrumentation will be presented. The MELODIE device is now ready and irradiation should start in OSIRIS reactor this spring. (authors)

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Assessment of LWR spent fuel disposal options. Volume 2. System definition and assessments. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-01

    Volume 2, in addition to introducing and defining the requirements for the study, contains the detailed evaluation of the study cases in each of six assessment areas, and describes and evaluates a number of study variations.

  1. Generic study on the relation between contamination if primary coolants and occupational radiation exposure in nuclear power plants with PWR. Final report; Generische Studie zum Zusammenhang zwischen Kontamination von Primaerkreislaufmedien und beruflicher Strahlenexposition bei Kernkraftwerken mit Druckwasserreaktor. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Artmann, Andreas; Bruhn, Gerd; Schneider, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany); Strub, Erik [Koeln Univ. (Germany)

    2016-01-15

    A generic model for the primary cooling system contamination in pressurized water reactors and the resulting radiological consequences has been developed. The functional capability was demonstrated by means of three examples concerning manipulation procedures during revision outages. Activities at the main reactor coolant pumps were studied and the influence of the coolant contamination on the resulting dose rates and collective doses were calculated. The effect of a Co-90 hot spot in a more remote area on the radiation exposure during the specific action at the reactor pumps was considered.

  2. Identification of the impacts of maintenance and testing upon the safety of LWR power plants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Husseiny, A. A.; Sabri, Z. A.; Turnage, J. J.

    1980-04-01

    The present study was designed to identify the impact of maintenance and testing (M and T) upon the safety of LWR power plants. The study involved data extraction from various sources reporting safety-related and operation-related nuclear power plant experience. Primary sources reviewed, including Licensee Event Reports (LER's) submitted to the NRC, revealed that only ten percent of events reported could be identified as M and T problems. The collected data were collated in a manner that would allow identification of principal types of problems which are associated with the performance of M and T tasks in LWR power plants. Frequencies of occurrence of events and their general endemic nature were analyzed using data clustering and pattern recognition techniques, as well as chi-square analyses for sparse contingency tables. The results of these analyses identified seven major categories of M and T error modes which were related to individual facilities and reactor type. Data review indicated that few M and T problems were directly related to procedural inadequacies, with the majority of events being attributable to human error.

  3. Proposal and analysis of the benchmark problem suite for reactor physics study of LWR next generation fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-10-01

    In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has established the Working Party on Reactor Physics for LWR Next Generation Fuels. The next generation fuels mean the ones aiming for further extended burn-up such as 70 GWd/t over the current design. The Working Party has proposed six benchmark problems, which consists of pin-cell, PWR fuel assembly and BWR fuel assembly geometries loaded with uranium and MOX fuels, respectively. The specifications of the benchmark problem neglect some of the current limitations such as 5 wt% {sup 235}U to achieve the above-mentioned target. Eleven organizations in the Working Party have carried out the analyses of the benchmark problems. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified. In this report, details of the benchmark problems, result by each organization, and their comparisons are presented. (author)

  4. 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.

  5. 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.

  6. 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.)

  7. 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.

  8. 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)

  9. Analysis of triso packing fraction and fissile material to DB-MHR using LWR reprocessed fuel

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R., E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: maritzargual@gmail.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    Gas-cooled and graphite-moderated reactor is being considered the next generation of nuclear power plants because of its characteristic to operate with reprocessed fuel. The typical fuel element consists of a hexagonal block with coolant and fuel channels. The fuel pin is manufactured into compacted ceramic-coated particles (TRISO) which are used to achieve both a high burnup and a high degree of passive safety. This work uses the MCNPX 2.6.0 to simulate the active core of Deep Burn Modular Helium Reactor (DB-MHR) employing PWR (Pressurized Water Reactor) reprocessed fuel. However, before a complete study of DB-MHR fuel cycle and recharge, it is necessary to evaluate the neutronic parameters to some values of TRISO Packing Fractions (PF) and Fissile Material (FM). Each PF and FM combination would generate the best behaviour of neutronic parameters. Therefore, this study configures several PF and FM combinations considering the heterogeneity of TRISO layers and lattice. The results present the best combination of PF and FM values according with the more appropriated behaviour of the neutronic parameters during the burnup. In this way, the optimized combination can be used to future works of MHR fuel cycle and recharge. (author)

  10. 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.

  11. 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.

  12. 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.

  13. Performance Evaluation of AI2O3/Water Nanofluid as Coolant in a Double-Tube Heat Exchanger Flowing under a Turbulent Flow Regime

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2012-01-01

    Full Text Available Nanofluids are expected to be a promising coolant candidate in chemical processes for water waste remediation and heat transfer system size reduction. This paper focuses on the potential mass flowrate reduction in exchanger with a given heat exchange capacity using nanofluids. Al2O3 nanoparticles with diameters of 7 nm dispersed in water with volume concentrations up to 2% are selected as a coolant, and their performance in a horizontal double-tube counterflow heat exchanger under turbulent flow conditions is numerically studied. The results show that the flowrate of nanofluid coolant decreases with the increase of concentration of nanoparticles in the exchanger with a given heat exchange capacity. The mass flowrate of the nanofluid at a volume concentration of 2 vol.% is approximately 24.5% lower than that of pure water (base fluid for given conditions. For the pressure drop, the results show that the pressure drop of nanofluid is slightly higher than water and increases with increase of volume concentrations. In addition, the reduction of wall temperature and heat transfer area is estimated.

  14. ROSA-III base test series for a large break loss-of-coolant accident in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tasaka, K.; Abe, N.; Anoda, Y.; Koizumi, Y.; Shiba, M.

    1982-05-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. It is confirmed from the experimental results obtained so far that the ROSA-III test facility can simulate major aspects of a BWR LOCA, such as boiling transition by lowering of the mixture level in the core, rewetting by the lower plenum flashing, and final quenching by the ECCS. The overall agreement between the calculated results by the RELAP5/ MOD0 code and the experimental results is good; however, the calculated lower plenum flashing rewetted the whole core and the calculated cladding temperature considerably underpredicts the measured value at the upper part of the core.

  15. 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.

  16. 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.

  17. 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.

  18. 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...

  19. 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)

  20. 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.

  1. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Science.gov (United States)

    Flaspoehler, Timothy; Petrovic, Bojan

    2016-02-01

    One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV). While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR) Power Plant, including DPA (displacements per atom) and fast neutron fluence (>1 MeV and >0.1MeV). I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling) methodology to bias a fixed-source MC (Monte Carlo) simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  2. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR

    Directory of Open Access Journals (Sweden)

    Flaspoehler Timothy

    2016-01-01

    Full Text Available One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV. While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR Power Plant, including DPA (displacements per atom and fast neutron fluence (>1 MeV and >0.1MeV. I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling methodology to bias a fixed-source MC (Monte Carlo simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  3. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2E, Physical descriptions of LWR nonfuel assembly hardware, Appendix 2F, User's guide to the LWR nonfuel assembly data base

    Energy Technology Data Exchange (ETDEWEB)

    None

    1987-12-01

    This appendix includes a two to three page Physical Description report for each Non-fuel Assembly (NFA) Hardware item identified from the current data. Information was obtained via subcontracts with these NFA hardware vendors: Babcock and Wildox, Combustion Engineering and Westinghouse. Data for some NFA hardware are not available. For such hardware, the information shown in this report was obtained from the open literature. Efforts to obtain additional information are continuing. NFA hardware can be grouped into six categories: BWR Channels, Control Elements, Guide Tube Plugs/Orifice Rods, Instrumentation, Neutron Poisons, and Neutron Sources. This appendix lists Physical Description reports alphabetically by vendor within each category. Individual Physical Description reports can be generated interactively through the menu-driven LWR Non-Fuel Assembly Hardware Data Base system. These reports can be viewed on the screen, directed to a printer, or saved in a text file for later use. Special reports and compilations of specific data items can be produced on request.

  4. 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.

  5. Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Merckx, K.R.; Holm, J.S.

    1981-01-01

    This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.

  6. 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)

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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...

  15. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  16. Large-break loss-of-coolant accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.G.; Sills, H.E.; Langman, V.J.; Boyack, B. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    The Advanced Candu Reactor (ACR) is an evolutionary advancement of the current Candu-6 reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table. (authors)

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 – Aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Howe, Kerry J., E-mail: howe@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Mitchell, Lana, E-mail: lmitchell@alionscience.com [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kim, Seung-Jun, E-mail: skim@lanl.gov [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Blandford, Edward D., E-mail: edb@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kee, Ernest J., E-mail: erniekee@gmail.com [South Texas Project Nuclear Operating Company, P.O. Box 270, Wadsworth, TX 77483 (United States)

    2015-10-15

    Highlights: • Trisodium phosphate (TSP) causes aluminum corrosion to cease after 24 h of exposure. • Chloride, iron, and copper have a minimal effect on the rate of aluminum corrosion when TSP is present. • Zinc can reduce the rate of aluminum corrosion when TSP is present. • Aluminum occasionally precipitates at concentrations lower than the calculated solubility for Al(OH){sub 3}. • Corrosion and solubility equations can be used to calculate the solids generated during a LOCA. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum from metallic aluminum surfaces under conditions representative of the containment pool following a postulated loss of coolant accident at a nuclear power generating facility. The experiments showed that TSP is capable of passivating the aluminum surface and preventing continued corrosion after about 24 h at the conditions tested. A correlation that describes the rate of corrosion including the passivation effect was developed from the bench experiments and validated with a separate set of experiments from a different test system. The saturation concentration of aluminum was shown to be well described by the solubility of amorphous aluminum hydroxide for the majority of cases, but instances have been observed when aluminum precipitates at concentrations lower than the calculated aluminum hydroxide solubility. Based on the experimental data and previous literature, an equation was developed to calculate the saturation concentration of aluminum as a function of pH and temperature under conditions representative of a loss of coolant accident (LOCA) in a TSP-buffered pressurized water reactor (PWR) containment. The corrosion equation and precipitation equation can be used in concert with each other to calculate the quantity of solids that would form as a function of time during a LOCA if the temperature and pH profiles were known.

  2. Estimate of coolant flow in assemblies of a natural circulation BWR applying and equivalent electric model; Estimacion del flujo de refrigerante en los ensambles de un BWR de circulacion natural aplicando un modelo electrico equivalente

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)], e-mail: julfi_jg@yahoo.com.mx

    2009-10-15

    The present work exposes the design and implementation of an advanced controller that it allows to estimate the coolant flow in fuel assemblies of a natural circulation BWR in real time. the complete development of this study is part of a doctoral project in course. In this work the construction of optimal controller is shown that allows to estimate the coolant flows in reactor and its operation applied to an equivalent electric model to natural circulation ESBWR. The controller design that allows the completely automatic starter of natural circulation reactor, required of a variables estimator not meter directly of nuclear power plant and use of local distributions estimates of coolant flow, (this controller type at the moment is utilized in the A BWR and several BWR in operation in Japan). The construction of estimator controller is mathematically based in the theory referring to Kalman filter, whose algorithm provides an advanced control of system. To prove the estimator operation was developed a simplified model that reproduces the basic dynamic of coolant flowing in the ESBWR, a practice way and very interesting of representing this phenomenon is by means the use of an equivalent electric model, which was developed starting from analogies that there is among the relation that keep the pressure differences with the mass flow and differences of electric potential with electric current. A detailed analysis of equivalence among models will be presented in a later article. (Author)

  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 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

  5. Investigation of the melt-down behaviour of massive radial core enclosures during LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Hering, W.; Sengpiel, W.; Messainguiral, C. [CEA Centre d' Etudes Nucleaires de Cadarache, 13 - Saint-Paul-les-Durance (France). DRN

    2000-11-01

    At the Institute for Reactor Safety (IRS) of the Forschungszentrum Karlsruhe (FZK) accident analyses were performed for the projected European pressurised water reactor (EPR) up to 1999 using the best estimate severe core damage code SCDAP/RELAP5 (S/R5). From various scenarios investigated with S/R5 the loss-of-offsite power (LOOP) and the 46 cm{sup 2} small break loss of coolant accident (SBLOCA) were selected to be discussed here in some detail. To simulate the heavy reflector (HR) and core barrel (CB) behaviour beyond the capabilities of S/R5 mod 3.2 a detailed stand alone analytical tool (LOWCOR2) was developed and used to determine the time of HR melting, its axial position, the melting velocity and the melt mass. Furthermore, results of MELCOR calculations performed at Siemens/KWU were used for the SBLOCA scenario. The analyses were extended by a feasibility study to find out whether ICARE2 and the commercial FEM code FIDAP are applicable. The axial position of HR and CB melt through strongly depends on the scenario an ranges between 1.0 m and 2.5 m core elevation. The time period to melt down the HR inner edges lasts up to 17 min and a complete melt through of HR and CB is in the order of magnitude of one hour. At melt through time LOWCOR2 calculated a molten steel mass between 10 Mg and 32 Mg and a melt relocation rate of 35 kg/s along the HR inner surface into the core cavity. (orig.) [German] Am Institut fuer Reaktorsicherheit (IRS) des Forschungszentrums Karlsruhe (FZK) wurden bis 1999 Unfallanalysen fuer den projektierten Europaeischen Druckwasser Reaktor (EPR) mit dem ''best estimate'' Kernschmelzcode SCDAP/RELAP5 (S/R5) durchgefuehrt. Von den verschiedenen mit S/R5 untersuchten Szenarien wurden der ''Ausfall der Wechselspannungsnetze'' (LOOP) und das kleine Leck (46 cm{sup 2}) im kalten Strang der Hauptkuehlmittelleitung (SBLOCA) fuer eine ausfuehrlichere Diskussion ausgewaehlt. Um das Verhalten des &apos

  6. Investigation of the Compatibility Between ADS Target Material With Coolant

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    At present, the program of the compatibility study on ADS project of China Institute of Atomic Energy is focused on the compatibility tests for the tungsten with water and sodium. An ADS verification facility is proposed in next phase of ADS project, an existent swimming pool reactor will be repack as the subcritical reactor system, and tungsten will be used as the target. On the other hand, our CEFR is being constructed now, it may be one of the options as the subcritical reactor system of ADS, thus, it is necessary to understand the compatibility characteristics of tungsten with sodium and water.

  7. Investigation of radioiodine behavior under accident conditions in LWR. Final report; Untersuchungen des Verhaltens von Radioiod unter Stoerfallbedingungen in LWR in kleintechnischen Versuchsanlagen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Panovsky, W.; Alm, M.; Eckelmann, U.; Jaspers, V.; Rentsch, J.; Vinzelberg, T.; Tuerpe, W.; Stark, M.

    1993-12-31

    Starting from safety considerations for WWER-type reactors main aspects of iodine chemistry are represented. By means of the use of application-oriented measuring techniques, existing equipment systems and their adaption to the requirements, investigations of the retention capacity of wet condensation systems for radioiodine were carried out. After analytical pre-tests (AAS, gaschromatography, laserspectrometry) - experiments were realized relating to both iodine absorption and iodine revolatilization. The absorption levels determined are between 92,5 and 99,9%, the iodine entrainment with revolatilization from the wet-condensation model is in the range between 0,003 and 0,0004%. (orig./HP) [Deutsch] Ausgehend von Sicherheitsbetrachtungen fuer WWER-Reaktoren werden die Hauptaspekte der Iodchemie dargelegt. Durch Nutzung anwendungsorientierter Messtechniken, vorhandener Geraetesysteme und deren Anpassung an die Erfordernisse wurden Untersuchungen zum Rueckhaltevermoegne von Nasskondensationssystemen fuer Radioiod angestellt. Nach analytischen Vorversuchen (AAS, Gaschromatographie, Laserspektrometrie) wurden Experimente zur Iodabsorption zur Iodwiederfreisetzung durchgefuehrt. Die ermittelten Absorptionsgrade lagen zwischen 92,5 und 99,9%, der Iodaustrag bei der Wiederfreisetzung aus dem Nasskondensationsmodell bewegt sich im Bereich von 0,003 und 0,0004%. (orig./HP)

  8. 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.

  9. 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.

  10. 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.

  11. Development of Reactor Coolant Pump for APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Sang-Youn; Chu, Sung-Min; Chang, Jin-Young [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

    2015-10-15

    The development was focused on the performance requirements for APR1400 and to achieve the goals of the safety, reliability and adaptability for APR1400 system design. In addition, APR1400 RCP design was customized considering convenience of installation, operation and maintainability. This paper describes the details of the development process, improved design feature and type test results. Based on development of core technology of RCP, DOOSAN supplies the localized and improved APR1400 RCP to Shin-Hanul 1 and 2 Project. This would be good experience that the RCP core technology can break foreign monopoly in supplying the domestic nuclear industry. Also, there expect APR1400 RCP can be sustainable revenue models in nuclear industry. Moreover, development of RCP will be a catalyst to enhance design capacity for equipment and system of nuclear power plant as well as evaluation and verification skills of Korean nuclear industry.

  12. Experimental Results on Pouring and Underwater Liquid Melt Spreading and Energetic Melt-coolant Interaction

    OpenAIRE

    Konovalenko, Alexander; Karbojian, Aram; Kudinov, Pavel

    2012-01-01

    In a hypothetical light water reactor (LWR) core-melt accident with corium release from the reactor  vessel,  the  ultimate  containment  integrity  is  contingent  on  coolability  of  the decay-heated core debris. Pouring of melt into a pool of water located in the reactor cavity is considered in several designs of existing and new LWRs  as a part of severe accident (SA) management strategies. At certain conditions of melt release into the pool (e.g. large ratio of the  vessel  breach  size...

  13. 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.

  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. 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.

  16. 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

  17. 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.

  18. Improved Traps for Removing Gases From Coolant Liquids

    Science.gov (United States)

    Holladay, John; Ritchie, Stephen

    2006-01-01

    Two documents discuss improvements in traps for removing noncondensable gases (e.g., air) from heat-transfer liquids (e.g., water) in spacecraft cooling systems. Noncondensable gases must be removed because they can interfere with operation. A typical trap includes a cylindrical hydrophobic membrane inside a cylindrical hydrophilic membrane, all surrounded by an outer cylindrical impermeable shell. The input mixture of gas bubbles and liquid flows into the annular volume between the membranes. Bubbles pass into the central hollow of the hydrophobic membrane and are vented. The liquid flows outward through the hydrophilic membrane and is recirculated.

  19. 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.

  20. 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.

  1. 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)

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Advanced Thermal Storage for Central Receivers with Supercritical Coolants

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Bruce D.

    2010-06-15

    The principal objective of the study is to determine if supercritical heat transport fluids in a central receiver power plant, in combination with ceramic thermocline storage systems, offer a reduction in levelized energy cost over a baseline nitrate salt concept. The baseline concept uses a nitrate salt receiver, two-tank (hot and cold) nitrate salt thermal storage, and a subcritical Rankine cycle. A total of 6 plant designs were analyzed, as follows: Plant Designation Receiver Fluid Thermal Storage Rankine Cycle Subcritical nitrate salt Nitrate salt Two tank nitrate salt Subcritical Supercritical nitrate salt Nitrate salt Two tank nitrate salt Supercritical Low temperature H2O Supercritical H2O Two tank nitrate salt Supercritical High temperature H2O Supercritical H2O Packed bed thermocline Supercritical Low temperature CO2 Supercritical CO2 Two tank nitrate salt Supercritical High temperature CO2 Supercritical CO2 Packed bed thermocline Supercritical Several conclusions have been drawn from the results of the study, as follows: 1) The use of supercritical H2O as the heat transport fluid in a packed bed thermocline is likely not a practical approach. The specific heat of the fluid is a strong function of the temperatures at values near 400 °C, and the temperature profile in the bed during a charging cycle is markedly different than the profile during a discharging cycle. 2) The use of supercritical CO2 as the heat transport fluid in a packed bed thermocline is judged to be technically feasible. Nonetheless, the high operating pressures for the supercritical fluid require the use of pressure vessels to contain the storage inventory. The unit cost of the two-tank nitrate salt system is approximately $24/kWht, while the unit cost of the high pressure thermocline system is nominally 10 times as high. 3) For the supercritical fluids, the outer crown temperatures of the receiver tubes are in the range of 700 to 800 °C. At temperatures of 700 °C and above

  7. 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.

  8. 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.

  9. 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)

  10. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  11. Assessment of alkali metal coolants for the ITER blanket

    Science.gov (United States)

    Natesan, K.; Reed, C. B.; Mattas, R. F.

    1994-06-01

    The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The blanket comparison and selection study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper addresses the thermodynamics of interactions between the liquid metals (e.g., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data are used to assess the long-term performance of the first wall in a liquid metal environment. Other key issues include development of electrical insulator coatings on the first-wall structural material to MHD pressure drop, and tritium permeation/inventory in self-cooled and indirectly cooled concepts. Acceptable types of coatings (based on their chemical compatibility and physical properties) are identified, and surface-modification avenues to achieve these coatings on the first wall are discussed. The assessment examines the extent of our knowledge on structural materials performance in liquid metals and identifies needed research and development in several of the areas in order to establish performance envelopes for the first wall in a liquid-metal environment.

  12. Thermochemical prediction of chemical form distributions of fission products in LWR oxide fuels irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-09-01

    Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. In addition, the effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that approximately 13 mole% of the total amount of Cs compounds exists as CsI and virtually remaining fraction as Cs{sub 2}MoO{sub 4} under the operation condition of LHR below 400 W/cm. On the other hand, when LHR is beyond 400 W/cm under the transient operation condition, its distribution did not change so much from the one under normal operation condition. (author)

  13. 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

  14. Thermal management systems and methods

    Science.gov (United States)

    Gering, Kevin L.; Haefner, Daryl R.

    2006-12-12

    A thermal management system for a vehicle includes a heat exchanger having a thermal energy storage material provided therein, a first coolant loop thermally coupled to an electrochemical storage device located within the first coolant loop and to the heat exchanger, and a second coolant loop thermally coupled to the heat exchanger. The first and second coolant loops are configured to carry distinct thermal energy transfer media. The thermal management system also includes an interface configured to facilitate transfer of heat generated by an internal combustion engine to the heat exchanger via the second coolant loop in order to selectively deliver the heat to the electrochemical storage device. Thermal management methods are also provided.

  15. Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

    Science.gov (United States)

    Proskuryakov, K. N.; Fedorov, A. I.; Zaporozhets, M. V.

    2015-08-01

    The accident at the Japanese Fukushima Daiichi nuclear power plant (NPP) caused by an earthquake showed the need of taking further efforts aimed at improving the design and engineering solutions for ensuring seismic resistance of NPPs with due regard to mutual influence of the dynamic processes occurring in the NPP building structures and process systems. Resonance interaction between the vibrations of NPP equipment and coolant pressure pulsations leads to an abnormal growth of dynamic stresses in structural materials, accelerated exhaustion of equipment service life, and increased number of sudden equipment failures. The article presents the results from a combined calculation-theoretical and experimental substantiation of mutual amplification of two kinds of external periodic loads caused by rotation of the reactor coolant pump (RCP) rotor and an earthquake. The data of vibration measurements at an NPP are presented, which confirm the predicted multiple amplification of vibrations in the steam generator and RCP at a certain combination of coolant thermal-hydraulic parameters. It is shown that the vibration frequencies of the main equipment may fall in the frequency band corresponding to the maximal values in the envelope response spectra constructed on the basis of floor accelerograms. The article presents the results from prediction of conditions under which vibroacoustic resonances with external periodic loads take place, which confirm the occurrence of additional earthquake-induced multiple growth of pressure pulsation intensity in the steam generator at the 8.3 Hz frequency and additional multiple growth of vibrations of the RCP and the steam generator cold header at the 16.6 Hz frequency. It is shown that at the elastic wave frequency equal to 8.3 Hz in the coolant, resonance occurs with the frequency of forced vibrations caused by the rotation of the RCP rotor. A conclusion is drawn about the possibility of exceeding the design level of equipment vibrations

  16. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  17. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A. E. [Oak Ridge National Lab., TN (United States); Tennessee Univ., Knoxville, TN (United States); Cheng, L. Y. [Brookhaven National Lab., Upton, NY (United States); Dimenna, R. A. [Westinghouse Savannah River Co., Aiken, SC (United States); Griffith, P. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Wilson, G. E. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis.

  18. 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.

  19. In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.

    1980-01-01

    In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, which has been observed in all tests with previously irradiated rods, needs further investigation.

  20. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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)

  9. Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2014

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-07-01

    The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet, and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security.

  10. 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.

  11. Sensitivity Analysis of Core Damage from Reactor Coolant Pump Seal Leakage during Extended Loss of All AC Power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Da Hee; Kim, Min Gi; Lee, Kyung Jin; Hwang, Su hyun; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Yoon, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    In this study, in order to comprehend the Fukushima accident, the sensitivity analysis was performed to analyze the behavior of Reactor Coolant System (RCS) during ELAP using the RELAP5/MOD3.3 code. The Fukushima accident was caused by tsunami resulted in Station Black Out (SBO) followed by the reactor core melt-down and release of radioactive materials. After the accident, the equipment and strategies for the Extended Loss of All AC Power (ELAP) were recommended strongly. In this analysis, sensitivity studies for the RCP seal failure of the OPR1000 type NPP were performed by using RELAP5/MOD3.3 code. Six cases with different leakage rate of RCP seal were studied for ELAP with operator action or not. The main findings are summarized as follows: (1) Without the operator action, the core uncovery time is determined by the leakage rate of RCP seal. When the leakage rate per RCP seal are 5 gpm, 50 gpm, and 300 gpm respectively, the core uncovery time are 1.62 hr, 1.58 hr, and 1.29 hr respectively. Namely, If the leakage rate of RCP seal was much bigger, the uncover time of core would be shorter. (2) In case that the cooling by SG secondary side was performed using the TDAFP and SG ADV, the core uncovery time was significantly extended.

  12. Development of advanced techniques for life management and inspection of advanced heavy water reactor (AWHR) coolant channel components

    Energy Technology Data Exchange (ETDEWEB)

    Madhusoodanan, K.; Sinha, S.K.; Kumar, K.; Shyam, T.V.; Panwar, S.; Sharma, B.S.V.G. [Bhabha Atomic Research Centre, Reactor Engineering Div., Trombay, Mumbai (India); Sinha, R. K. [Bhabha Atomic Research Centre, Reactor Design and Development Group., Trombay, Mumbai (India)

    2011-07-01

    Operating life of pressure tubes of Pressurized Heavy Water Reactor (PHWR) is limited due to the presence of various issues associated with the material like hydrogen pick up, delayed hydride cracking, axial elongation and increase in diameter due to irradiation creep and growth. Periodic monitoring of the health of the pressure tube under in-situ conditions is essential to ensure the safe operation of the reactor. New designs of reactor call for innovative design philosophy, modification in fabrication route of pressure tube, development of reactor specific tools, both analytical and hardware for assessing the fitness for service of the pressure tube. Feedback from existing reactors has enhanced the understanding about life limiting parameters. This paper gives an insight into the life limiting issues associated with pressure tube and the efforts pursued for development of life management techniques for coolant channel of Advanced Heavy Water Reactor (AHWR) designed in India. The tools and techniques for in-situ property/hydrogen measurement, pulsed eddy current technique for zirconium alloy in-homogeneity characterization, horizontal shear wave EMAT system for dissimilar metal weld inspection, sliver sampling of vertical channel etc. are elaborated in the paper. (author)

  13. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

    Science.gov (United States)

    Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-01

    This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

  14. 基于相依竞争失效模型的主泵多状态可靠性分析%Multi-state Reliability for Coolant Pump Based on Dependent Competitive Failure Model

    Institute of Scientific and Technical Information of China (English)

    尚彦龙; 蔡琦; 赵新文; 陈玲

    2013-01-01

    基于核动力主泵运行环境和性能退化机理,考虑自身振动和外部冲击对其性能退化的影响,建立了主泵冲击与退化相依竞争失效过程的可靠度模型。采用该模型计算了考虑性能退化的主泵在振动和外部冲击条件下的退化状态概率和可靠度,为基于使用环境的核动力主泵的多状态可靠性分析提供了一种有效的分析途径。分析结果可为设计变更和维修优化提供决策依据。%By taking into account the effect of degradation due to internal vibration and external shocks , and based on service environment and degradation mechanism of nuclear power plant coolant pump ,a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation .Using this model ,degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump .It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment .The results can provide a decision making basis for design changing and maintenance optimization .

  15. 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.

  16. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

  17. Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    1977-03-01

    This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. Support of the launching of motor car air conditioning systems with the coolant CO{sub 2} (R744). Test bench measurements and practical trials; Unterstuetzung der Markteinfuehrung von Pkw-Klimaanlagen mit dem Kaeltemittel CO{sub 2} (R744). Pruefstandsmessungen und Praxistest

    Energy Technology Data Exchange (ETDEWEB)

    Lemke, Nicholas; Mildenberger, Julia [Technische Univ. Braunschweig (Germany); Graz, Martin [Obrist Engineering GmbH, Lustenau (Austria)

    2011-10-15

    In the research project two passenger car air-conditioning systems were analyzed with regard to cooling capacity and efficiency. The results were compared with one another. The first system was a standard air-conditioning unit using R134a as a refrigerant. As a second system a CO{sub 2} (R744) prototype HVAC unit was used. Both units were investigated on one hand installed in a car on a dynamometer by Obrist Engineering GmbH and on the other hand installed in a calorimetric test rig by Technische Universitaet Braunschweig, Institut fuer Thermodynamik. While the tests in the calorimetric test rig showed comparable efficiencies and cooling capacities for both setups, consumption advantages were determined for the R744- air-conditioning unit installed in the vehicle by the company Obrist. With CO{sub 2} (R744) as a refrigerant for mobile air-conditioning systems an environmental friendly solution is available. (orig.)

  9. System and method for conditioning intake air to an internal combustion engine

    Energy Technology Data Exchange (ETDEWEB)

    Sellnau, Mark C.

    2015-08-04

    A system for conditioning the intake air to an internal combustion engine includes a means to boost the pressure of the intake air to the engine and a liquid cooled charge air cooler disposed between the output of the boost means and the charge air intake of the engine. Valves in the coolant system can be actuated so as to define a first configuration in which engine cooling is performed by coolant circulating in a first coolant loop at one temperature, and charge air cooling is performed by coolant flowing in a second coolant loop at a lower temperature. The valves can be actuated so as to define a second configuration in which coolant that has flowed through the engine can be routed through the charge air cooler. The temperature of intake air to the engine can be controlled over a wide range of engine operation.

  10. 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.

  11. Application of an intermediate LWR for electricity production and hot-water district heating

    Energy Technology Data Exchange (ETDEWEB)

    1983-05-01

    The objective of the study is to evaluate the technical and economic feasibility of a 400 MWe Consolidated Nuclear Steam System (CNSS) for supplying district heat to the Minneapolis/St. Paul area. A total of three CNSS reactor sites, located various distances from the Minneapolis-St. Paul area load center, are evaluated. The distance from the load center is determined by the credited safety features of the plant design. Each site is also evaluated for three different hot water supply/return temperatures providing a total of nine CNSS study cases. The cost of district heat delivered to the load center is determined for each case.

  12. Contribution to the optimization of the chemical and radiochemical purification of pressurized water nuclear power plants primary coolant; Contribution a l'optimisation de la purification chimique et radiochimique du fluide primaire des centrales nucleaires a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Elain, L

    2004-12-15

    The primary coolant of pressurised water reactors is permanently purified thanks to a device, composed of filters and the demineralizers furnished with ion exchange resins (IER), located in the chemical and volume control system (CVCS). The study of the retention mechanisms of the radio-contaminants by the IER implies, initially, to know the speciation of the primary coolant percolant through the demineralizers. Calculations of theoretical speciation of the primary coolant were carried out on the basis of known composition of the primary coolant and thanks to the use of an adapted chemical speciation code. A complementary study, dedicated to silver behaviour, considered badly extracted, suggests metallic aggregates existence generated by the radiolytic reduction of the Ag{sup +} ions. An analysis of the purification curves of the elements Ni, Fe, Co, Cr, Mn, Sb and their principal radionuclides, relating to the cold shutdown of Fessenheim 1-cycle 20 and Tricastin 2-cycle 21, was carried out, in the light of a model based on the concept of a coupling well term - source term. Then, a thermodynamic modelling of ion exchange phenomena in column was established. The formation of the permutation front and the enrichment zones planned was validated by frontal analysis experiments of synthetic fluids (mixtures of Ni(B(OH){sub 4}){sub 2}, LiB(OH){sub 4} and AgB(OH){sub 4} in medium B(OH){sub 3})), and of real fluid during the putting into service of the device mini-CVCS at the time of Tricastin 2 cold shutdown. New tools are thus proposed, opening the way with an optimised management of demineralizers and a more complete interpretation of the available experience feedback. (author)

  13. 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.

  14. Plutonium rock-like fuel LWR nuclear characteristics and transient behavior in accidents

    Energy Technology Data Exchange (ETDEWEB)

    Akie, Hiroshi; Anoda, Yoshinari; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yamaguchi, Chouichi; Sugo, Yukihiro

    1998-03-01

    For the disposition of excess plutonium, rock-like oxide (ROX) fuel systems based on zirconia (ZrO{sub 2}) or thoria (ThO{sub 2}) have been studied. Safety analysis of ROX fueled PWR showed it is necessary to increase Doppler reactivity coefficient and to reduce power peaking factor of zirconia type ROX (Zr-ROX) fueled core. For these improvements, Zr-ROX fuel composition was modified by considering additives of ThO{sub 2}, UO{sub 2} or Er{sub 2}O{sub 3}, and reducing Gd{sub 2}O{sub 3} content. As a result of the modification, comparable, transient behavior to UO{sub 2} fuel PWR was obtained with UO{sub 2}-Er{sub 2}O{sub 3} added Zr-ROX fuel, while the plutonium transmutation capability is slightly reduced. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  16. Fuel performance improvement program. Quarterly progress report, April--June 1978. [LWR

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1978-07-01

    The Fuel Performance Improvement Program has as its objective the identification and demonstration of fuel concepts with improved power ramp performance. Improved fuels are being sought to allow reduction or elimination of fuel related operating guidelines on nuclear power plants such that the fuel may be power maneuvered within the rates allowed by the system technical specifications. The program contains a combination of out-of-reactor studies, in-reactor experiments and in-reactor demonstrations. Fuel concepts initially being considered include annular pellets, cladding internally coated with graphite and packed-particle fuels. The performance capability of each concept will be compared to a reference fuel of contemporary pellet design by irradiations in the Halden Boiling Water Reactor. Fuel design and process development is being completed and fuel rod fabrication will begin for the Halden test rods and for the first series of in-reactor experiments. The in-reactor demonstrations are being performed in the Big Rock Point reactor to show that the concepts pose no undue risk to commercial operation. Additional concepts may be considered as the result of a state-of-the-technology review of fuel-cladding interaction and assessment of fuel concepts and the out-of-reactor studies. The results of the program will be used to establish the technical bases for design of fuels with improved power ramp performance.

  17. 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

  18. 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

  19. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  20. 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

  1. 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)

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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)

  13. 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.

  14. 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.

  15. Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; R. S. Sen; M. A. Pope; A. M. Ougouag

    2011-09-01

    The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO{sub 2} kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 {micro}mm kernel diameter, 100 {micro}mm buffer, 35 {micro}mm IPyC, 35 {micro}mm SiC, 40 {micro}mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, results in a 8.0 x 10{sup -2} failure probability. For a &apos

  16. 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.

  17. 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.

  18. Development and performance of a large-scale, transonic turbine blade cascade facility for aerodynamic studies of merging coolant-mainstream flows

    Science.gov (United States)

    Al-Sayeh, Amjad Isaaf

    1998-11-01

    A new, large scale, linear cascade facility of turbine blades has been developed for the experimental exploration of the aerodynamic aspects of film cooling technology. Primary interest is in the mixing of the ejected coolant with the mainstream, at both subsonic and supersonic mainstream Mach numbers at the cascade exit. In order to achieve a spatial resolution adequate for the exploration of details on the scale of the coolant ejection holes, the cascade dimensions were maximized, within the limitations of the air supply system. The cascade contains four blades (three passages) with 14.05 cm axial chord, 17.56 cm span and a design total turning angle of 130.6 degrees. Exit Mach numbers range from 0.6 to 1.5 and Reynolds numbers from 0.5 to 1.5 million. The air supply system capacity allows run times up to five minutes at maximum flow rates. A coolant supply system has been built to deliver mixtures of SFsb6 and air to simulate coolant/mainstream density ratios up to 2. The cascade contains several novel features. A full-perimeter bleed slot upstream of the blades is used to remove the approach boundary layer from all four walls, to improve the degree of two-dimensionality. The exit flow is bounded by two adjustable tailboards that are hinged at the trailing edges and actuated to set the exit flow direction according to the imposed pressure ratio. The boards are perforated and subjected to mass removal near the blades, to minimize the undesirable reflection of shocks and expansion waves. A probe actuator is incorporated that allows continuous positioning of probes in the exhaust stream, in both the streamwise and pitchwise directions. Diagnostic methods include extensive surface pressure taps on the approach and exhaust ducts and on the blade surfaces. The large size permitted as many as 19 taps on the trailing edge itself. Shadowgraph and schlieren are available. A three-prong wake probe has been constructed to simultaneously measure total and static pressures

  19. Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2017-01-01

    Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

  20. Control of Thermodynamical System with Input-Dependent State Delays

    DEFF Research Database (Denmark)

    Bendtsen, Jan Dimon; Krstic, Miroslav

    2013-01-01

    We consider control of a cooling system with several consumers that require cooling from a common source. The flow feeding coolant to the consumers can be controlled, but due to significant physical distances between the common source and the consumers, the coolant flow takes a non-negligible amo...