WorldWideScience

Sample records for water reactors flux

  1. Neutron fluxes in test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  2. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  3. Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

    2017-05-15

    Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

  4. Reactor flux calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lhuillier, D. [Commissariat à l' Énergie Atomique et aux Énergies Alternatives, Centre de Saclay, IRFU/SPhN, 91191 Gif-sur-Yvette (France)

    2013-02-15

    The status of the prediction of reactor anti-neutrino spectra is presented. The most accurate method is still the conversion of total β spectra of fissionning isotopes as measured at research reactors. Recent re-evaluations of the conversion process led to an increased predicted flux by few percent and were at the origin of the so-called reactor anomaly. The up to date predictions are presented with their main sources of error. Perspectives are given on the complementary ab-initio predictions and upcoming experimental cross-checks of the predicted spectrum shape.

  5. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  6. Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.

  7. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

    2017-06-15

    Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

  8. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  9. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  10. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  11. Monitoring Akkuyu Nuclear Reactor Using Antineutrino Flux Measurement

    OpenAIRE

    Ozturk, Sertac; Adiguzel, Aytul; Ozcan, V. Erkcan; Unel, Gokhan

    2016-01-01

    We present a simulation-based study for monitoring Akkuyu Nuclear Power Plant's activity using antineutrino flux originating from the reactor core. A water Cherenkov detector has been designed and optimization studies have been performed using Geant4 simulation toolkit. A first study for the design of a monitoring detector facility for Akkuyu Nuclear Power Plant is discussed in this paper.

  12. Dissecting Reactor Antineutrino Flux Calculations

    Science.gov (United States)

    Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

    2017-09-01

    Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

  13. High flux diffractometers on reactor neutron sources

    Science.gov (United States)

    Hewat, Alan W.

    2006-11-01

    Continuous neutron sources such as reactors can deliver a very high time-averaged flux to the sample using a relatively wide band of wavelengths, while still retaining good resolution. For example, the D20 diffractometer at ILL Grenoble, the world's highest flux neutron powder machine, can collect complete patterns at 100 ms intervals, and this has been important for the real time study of explosive SHS reactions. New very large 2D detectors, such as those recently installed on D2B and D19 at ILL, are up to an order of magnitude larger than previous designs, and will provide unmatched speed of data collection from very small samples, opening up new scientific perspectives for powder and single crystal diffraction. We will discuss future reactor based diffractometers designed for rapid data collection from small samples in special environments.

  14. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  15. CHIMNEY FOR BOILING WATER REACTOR

    Science.gov (United States)

    Petrick, M.

    1961-08-01

    A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

  16. The High Flux Beam Reactor at Brookhaven National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, S.M.

    1994-12-31

    Brookhaven National Laboratory`s High Flux Beam Reactor (HFBR) was built because of the need of the scientist to always want `more`. In the mid-50`s the Brookhaven Graphite reactor was churning away producing a number of new results when the current generation of scientists, led by Donald Hughes, realized the need for a high flux reactor and started down the political, scientific and engineering path that led to the BFBR. The effort was joined by a number of engineers and scientists among them, Chemick, Hastings, Kouts, and Hendrie, who came up with the novel design of the HFBR. The two innovative features that have been incorporated in nearly all other research reactors built since are: (i) an under moderated core arrangement which enables the thermal flux to peak outside the core region where beam tubes can be placed, and (ii) beam tubes that are tangential to the core which decrease the fast neutron background without affecting the thermal beam intensity. Construction began in the fall of 1961 and four years later, at a cost of $12 Million, criticality was achieved on Halloween Night, 1965. Thus began 30 years of scientific accomplishments.

  17. HEAVY WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

    Szilard, L.

    1958-04-29

    A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

  18. High-flux first-wall design for a small reversed-field pinch reactor

    Science.gov (United States)

    Cort, G. E.; Graham, A. L.; Christensen, K. E.

    To achieve the goal of a commercially economical fusion power reactor, small physical size and high power density should be combined with simplicity (minimized use of high technology systems). The Reversed-Field Pinch (RFP) is a magnetic confinement device that promises to meet these requirements with power densities comparable to those in existing fission power plants. To establish feasibility of such an RFP reactor, a practical design for a first wall capable of withstanding high levels of cyclic neutron wall loadings is needed. Associated with the neutron flux in the proposed RFP reactor is a time averaged heat flux of 4.5 MW/sq m with a conservatively estimated transient peak approximately twice the average value. The design for a modular first wall made from a high-strength copper alloy that will meet these requirements of cyclic thermal loading is presented. The heat removal from the wall is by subcooled water flowing in straight tubes at high linear velocities.

  19. Prospects for improved understanding of isotopic reactor antineutrino fluxes

    Science.gov (United States)

    Gebre, Y.; Littlejohn, B. R.; Surukuchi, P. T.

    2018-01-01

    Predictions of antineutrino fluxes produced by fission isotopes in a nuclear reactor have recently received increased scrutiny due to observed differences in predicted and measured inverse beta decay (IBD) yields, referred to as the "reactor antineutrino flux anomaly." In this paper, global fits are applied to existing IBD yield measurements to produce constraints on antineutrino production by individual plutonium and uranium fission isotopes. We find that fits including measurements from highly U 235 -enriched cores and fits including Daya Bay's new fuel evolution result produce discrepant best-fit IBD yields for U 235 and Pu 239 . This discrepancy can be alleviated in a global analysis of all data sets through simultaneous fitting of Pu 239 , U 235 , and U 238 yields. The measured IBD yield of U 238 in this analysis is (7.02 ±1.65 )×10-43 cm2/fission , nearly two standard deviations below existing predictions. Future hypothetical IBD yield measurements by short-baseline reactor experiments are examined to determine their possible impact on the global understanding of isotopic IBD yields. It is found that future improved short-baseline IBD yield measurements at both high-enriched and low-enriched cores can significantly improve constraints for U 235 , U 238 , and Pu 239 , providing comparable or superior precision to existing conversion- and summation-based antineutrino flux predictions. Systematic and experimental requirements for these future measurements are also investigated.

  20. High Flux Metallic Membranes for Hydrogen Recovery and Membrane Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buxbaum, Robert

    2010-06-30

    We made and tested over 250 new alloys for use as lower cost, higher flux hydrogen extraction membrane materials. Most of these were intermetallic, or contained significant intermetallic content, particularly based on B2 alloy compositions with at least one refractory component; B2 intermetallics resemble BCC alloys, in structure, but the atoms have relatively fixed positions, with one atom at the corners of the cube, the other at the centers. The target materals we were looking for would contain little or no expensive elements, no strongly toxic or radioactive elements, would have high flux to hydrogen, while being fabricable, brazable, and relatively immune to hydrogen embrittlement and corrosion in operation. The best combination of properties of the membrane materials we developed was, in my opinion, a Pd-coated membrane consisting of V -9 atomic % Pd. This material was relatively cheap, had 5 times the flux of Pd under the same pressure differential, was reasonably easy to fabricate and braze, and not bad in terms of embrittlement. Based on all these factors we project, about 1/3 the cost of Pd, on an area basis for a membrane designed to last 20 years, or 1/15 the cost on a flux basis. Alternatives to this membrane replaced significant fractions of the Pd with Ni and or Co. The cost for these membranes was lower, but so was the flux. We produced successful brazed products from the membrane materials, and made them into flat sheets. We tested, unsuccessfully, several means of fabricating thematerials into tubes, and eventually built a membrane reactor using a new, flat-plate design: a disc and doughnut arrangement, a design that seems well- suited to clean hydrogen production from coal. The membranes and reactor were tested successfully at Western Research. A larger equipment company (Chart Industries) produced similar results using a different flat-plate reactor design. Cost projections of the membrane are shown to be attractive.

  1. Very high flux research reactors based on particle fuels

    Energy Technology Data Exchange (ETDEWEB)

    Powell, J.R.; Takahashi, H.

    1985-01-01

    A new approach to high flux research reactors is described, the VHFR (Very High Flux Reactor). The VHFR fuel region(s) are packed beds of HTGR-type fuel particles through which coolant (e.g., D/sub 2/O) flows directly. The small particle diameter (typically on the order of 500 microns) results in very large surface areas for heat transfer (approx. 100 cm/sup 2//cm/sup 3/ of bed), high power densities (approx. 10 megawatts per liter), and minimal ..delta..T between fuel and coolant (approx. 10 K) VHFR designs are presented which achieve steady-state fluxes of approx. 2x10/sup 16/ n/cm/sup 2/sec. Deuterium/beryllium combinations give the highest flux levels. Critical mass is low, approx. 2 kg /sup 235/U for 20% enriched fuel. Refueling can be carried out continuously on-line, or in a batch process with a short daily shutdown. Fission product inventory is very low, approx. 100 to 300 grams, depending on design.

  2. SUPERHEATING IN A BOILING WATER REACTOR

    Science.gov (United States)

    Treshow, M.

    1960-05-31

    A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

  3. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  4. A conceptual high flux reactor design with scope for use in ADS ...

    Indian Academy of Sciences (India)

    Home; Journals; Pramana – Journal of Physics; Volume 68; Issue 2. A conceptual high flux reactor design with scope for use in ADS applications. Usha Pal V Jagannathan ... A 100 MWt reactor design has been conceived to support flux level of the order of 1015 n/cm2/s in selected flux trap zones. The physics design ...

  5. Cleaning natural water in the clarifier reactor

    Science.gov (United States)

    Skolubovich, Yuriy; Skolubovich, Aleksandr; Voitov, Evgeniy; Soppa, Mikhail; Chirkunov, Yuriy

    2017-10-01

    The problems of cleaning low turbidity high-color surface waters for drinking water supply are considered. A new design of the clarifier reactor is proposed, which increases the efficiency of water purification and at the same time reduces its operating costs. A detailed description of clarifier reactor design and its operation is given. The study results of the clarifier reactor operation in real conditions for the purification of low turbidity high-color waters are shown. Due to the weighted layer of dense loading use in a process of water purification, the structure productivity can be increased by 2-3 times in comparison with conventional clarifiers with suspended sediment. Using reagents for water purification, the clarifier reactor, due to the processes of contact coagulation, allows reducing the consumption of reagents up to 50%. Investigations of the clarifier reactor operation in technological schemes for various waters purification, including sewage, showed their effectiveness and prospects.

  6. Fabrication of control rods for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  7. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  8. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Prashant K [ORNL; Freels, James D [ORNL

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  9. High Flux Isotope Reactor system RELAP5 input model

    Energy Technology Data Exchange (ETDEWEB)

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  10. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  11. Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

  12. Response of the Gamma TIP Detectorsin a Nuclear Boiling Water Reactor

    OpenAIRE

    Fridström, Richard

    2010-01-01

    In order to monitor a nuclear boiling water reactor fixed and movable detectors are used, such as the neutron sensitive LPRM (Local Power Range Monitors) detectors and the gamma sensitive TIP (Traversing Incore Probe) detectors. These provide a mean to verify the predictions obtained from core simulators, which are used for planning and following up the reactor operation. The core simulators calculate e.g. the neutron flux and power distribution in the reactor core. The simulators can also si...

  13. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  14. Level 1 Tornado PRA for the High Flux Beam Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

  15. The development of ex-core neutron flux monitoring system for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S

    2004-12-01

    Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

  16. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  17. Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-07-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

  18. SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

    Science.gov (United States)

    Ransohoff, J.A.; Plawchan, J.D.

    1960-08-16

    A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

  19. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    Energy Technology Data Exchange (ETDEWEB)

    Austin, W.; Brinkley, D.

    2011-10-13

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  20. A conceptual high flux reactor design with scope for use in ADS ...

    Indian Academy of Sciences (India)

    Abstract. A 100 MWt reactor design has been conceived to support flux level of the order of 1015 n/cm2/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D2O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch ...

  1. Dissolution flowsheet for high flux isotope reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Foster, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  2. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  3. Light-water reactor accident classification

    Energy Technology Data Exchange (ETDEWEB)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art.

  4. 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.)

  5. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  6. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    Science.gov (United States)

    Treshow, M.

    1963-04-30

    This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

  7. Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

  8. Sterile neutrinos or flux uncertainties? — Status of the reactor anti-neutrino anomaly

    Science.gov (United States)

    Dentler, Mona; Hernández-Cabezudo, Álvaro; Kopp, Joachim; Maltoni, Michele; Schwetz, Thomas

    2017-11-01

    The ˜ 3 σ discrepancy between the predicted and observed reactor anti-neutrino flux, known as the reactor anti-neutrino anomaly, continues to intrigue. The recent discovery of an unexpected bump in the reactor anti-neutrino spectrum, as well as indications that the flux deficit is different for different fission isotopes seems to disfavour the explanation of the anomaly in terms of sterile neutrino oscillations. We critically review this conclusion in view of all available data on electron (anti)neutrino disappearance. We find that the sterile neutrino hypothesis cannot be rejected based on global data and is only mildly disfavored compared to an individual rescaling of neutrino fluxes from different fission isotopes. The main reason for this is the presence of spectral features in recent data from the NEOS and DANSS experiments. If state-of-the-art predictions for reactor fluxes are taken at face value, sterile neutrino oscillations allow a consistent description of global data with a significance close to 3 σ relative to the no-oscillation case. Even if reactor fluxes and spectra are left free in the fit, a 2 σ hint in favour of sterile neutrinos remains, with allowed parameter regions consistent with an explanation of the anomaly in terms of oscillations.

  9. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas, allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  10. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  11. Anti-neutrino flux in a research reactor for non-proliferation application

    Energy Technology Data Exchange (ETDEWEB)

    Khakshournia, Samad; Foroughi, Shokoufeh [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Atomic Energy Organization of Iran (AEOI)

    2017-11-15

    Owing to growing interest in the study of emitted antineutrinos from nuclear reactors to test the Atomic Energy Agency safeguards, antineutrino flux was studied in the Tehran Research Reactor (TRR) using ORIGEN code. According to our prediction, antineutrino rate was obtained 2.6 x 10{sup 17} (v{sub e}/sec) in the core No. 57F of the TRR. Calculations indicated that evolution of antineutrino flux was very slow with time and the performed refueling had not an observable effect on antineutrino flux curve for a 5 MW reactor with the conventional refueling program. It is seen that for non-proliferation applications the measurement of the contribution of {sup 239}Pu to the fission using an antineutrino detector is not viable in the TRR.

  12. FOEHN: The critical experiment for the Franco-German High Flux Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scharmer, K.; Eckert, H. G. [Kernforschungszentrum Karlsruhe GmbH (Germany)

    1991-01-01

    A critical experiment for the Franco-German High Flux Reactor was carried out in the French reactor EOLE (CEN Cadarache). The purpose of the experiment was to check the calculation methods in a realistic geometry and to measure effects that can only be calculated imprecisely (e.g. beam hole effects). The structure of the experiment and the measurement and calculation methods are described. A detailed comparison between theoretical and experimental results was performed. 30 refs., 105 figs.

  13. Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements.

    Science.gov (United States)

    Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-10-01

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. Copyright © 2015 Elsevier Ltd. All rights reserved.

  14. Boiling water neutronic reactor incorporating a process inherent safety design

    Science.gov (United States)

    Forsberg, C.W.

    1985-02-19

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  15. Apparatus for high flux photocatalytic pollution control using a rotating fluidized bed reactor

    Science.gov (United States)

    Tabatabaie-Raissi, Ali; Muradov, Nazim Z.; Martin, Eric

    2003-06-24

    An apparatus based on optimizing photoprocess energetics by decoupling of the process energy efficiency from the DRE for target contaminants. The technique is applicable to both low- and high-flux photoreactor design and scale-up. An apparatus for high-flux photocatalytic pollution control is based on the implementation of multifunctional metal oxide aerogels and other media in conjunction with a novel rotating fluidized particle bed reactor.

  16. High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

    2017-01-24

    Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

  17. Pressurized-water reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  18. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    Science.gov (United States)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  19. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  20. Neutron flux spectra in the FFTF In-Reactor Thimble

    Energy Technology Data Exchange (ETDEWEB)

    Wootan, D.W.; Dobbin, K.D.; Schmittroth, F.A.; Roberts, J.H.

    1982-12-01

    Neutron spectra measured in the FFTF In-Reactor Thimble (IRT) by proton recoil proportional counters, proton recoil emulsions, and passive dosimeters have been evaluated and compared with each other and with three-dimensionl diffusion theory calculated spectra for the purpose of validating the passive dosimeter spectrum adjustment technique. The least squares data adjustment code, FERRET, was used to combine measured reaction rates, calculated spectra, and dosimeter cross sections, resulting in adjusted spectra and cross sections with uncertainties and correlations that properly account for uncertainties and correlations on the input parameters.

  1. Catalytic membrane reactor for water and wastewater treatment

    Science.gov (United States)

    Heng, Samuel

    A double membrane reactor was fabricated and assessed for continuous treatment of water containing organic contaminants by ozonation. This innovative reactor consisted of a zeolite membrane prepared on the inner surface of a porous a-alumina support, which served as water selective extractor and active contactor, and a porous stainless membrane which was the ozone gas diffuser. The coupling of membrane separation and chemical oxidation was found to be highly beneficial to both processes. The total organic carbon (TOC) removal rate at the retentate was enhanced by up to 2.2 times, as compared to membrane ozonation. Simultaneously, clean water (membrane support, was shown to further enhance TOC degradation, permeated TOC concentration, permeate flux, and moreover, ozone yield. The achievements of this project included: (1) The development of a novel low-temperature zeolite membrane activation method that generates consistently high quality membranes (i.e. high reproducibility and fewer defects). (2) The demonstration that gamma-alumina and gamma-alumina supported catalysts do not have significant activity and that the TOC removal enhancement usually observed during catalytic ozonation was due primarily to the contribution of adsorption and metal leaching. Thermogravimetric analysis (TGA) and elemental analysis (EA) of the spent catalyst showed that, during catalytic ozonation, oxygenated by-products of increased adsorbability were concentrated onto the gamma-alumina contactor, and were subsequently degraded. (3) The development of a method for coating high surface area gamma-alumina layers onto the grains of zeolite membrane support used as the active membrane contactor.

  2. Reactor core and plant design concepts of the Canadian supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Bailey, J.; Rhodes, D.; Guzonas, D.; Hamilton, H.; Haque, Z.; Pencer, J.; Sartipi, A. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada is developing a 1200 MWe supercritical water-cooled reactor (SCWR), which has evolved from the well-established pressure-tube type CANDU{sup 1} reactor. This SCWR reactor concept, which is often referred to as the Canadian SCWR, uses supercritical water as a coolant, has a low-pressure heavy water moderator and a direct cycle for power production. The reactor concept incorporates advanced safety features, such as passive emergency core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce the core damage frequency beyond existing nuclear reactors. This paper presents a description of the Canadian SCWR core design concept, the integration of in-core and out-of-core components and the mechanical plant design concept. Supporting systems for reactor safety, reactor control and moderator cooling are also described. (author)

  3. High Heat Flux Testing of Tungsten PFCs for the Fusion Reactor Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Park, Seong Dae; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Tungsten divertor plasma facing components (PFCs) were fabricated by using hot isostatic-pressing (HIP) bonding and vacuum plasma spray method (VPS) for the fusion reactor development. Tungsten mockups were designed for the high heat flux test with the Korean high heat flux test facility - electron beam (KoHLT-EB), which was constructed for the performance qualification of various PFCs. KoHLT-EB was used to qualify the PFCs performances of each fusion research. The high heat load conditions were simulated by using a thermo-hydraulic analysis tool with the ANSYS-CFX code. A high heat flux testing was performed up to the thermal life-time of each mockup with the evaluated conditions in 1 ⁓ 5 MW/m{sup 2}. HIP-bonded testing mockups consisted of 2 mm pure tungsten PFCs and ferritic-martensitic steel (FMS) as structural materials, and VPS mockups are comprised with 1 ⁓ 2 mm coated tungsten layer and FMS. These tungsten mockups are the candidate for the blanket and divertor in DEMO or a fusion power reactor. For the thermal fatigue test, two types of tungsten mock-ups were fabricated by using HIP bonding and VPS method for the fusion reactor development. The verification of the cooling performance was tested under the operation condition of ITER and fusion reactor. After the completion of the preliminary mockup test and facility qualification, the high heat flux test facility assess the performance test for two type of plasma facing components.

  4. Measurement of Turbulent Fluxes of Swirling Flow in a Scaled Up Multi Inlet Vortex Reactor

    Science.gov (United States)

    Olsen, Michael; Hitimana, Emmanual; Hill, James; Fox, Rodney

    2017-11-01

    The multi-inlet vortex reactor (MIVR) has been developed for use in the FlashNanoprecipitation (FNP) process. The MIVR has four identical square inlets connected to a central cylindrical mixing chamber with one common outlet creating a highly turbulent swirling flow dominated by a strong vortex in the center. Efficient FNP requires rapid mixing within the MIVR. To investigate the mixing, instantaneous velocity and concentration fields were acquired using simultaneous stereoscopic particle image velocimetry and planar laser-induced fluorescence. The simultaneous velocity and concentration data were used to determine turbulent fluxes and spatial cross-correlations of velocity and concentration fluctuations. The measurements were performed for four inlet flow Reynolds numbers (3250, 4875, 6500, and 8125) and at three measurement planes within the reactor. A correlation between turbulent fluxes and vortex strength was found. For all Reynolds numbers, turbulent fluxes are maximum in the vortex dominated central region of the reactor and decay away from the vortex. Increasing Reynolds number increased turbulent fluxes and subsequently enhanced mixing. The mixing performance was confirmed by determining coefficients of concentration variance within the reactor.

  5. Improved Measurement of the Reactor Antineutrino Flux and Spectrum at Daya Bay

    CERN Document Server

    An, F P; Band, H R; Bishai, M; Blyth, S; Cao, D; Cao, G F; Cao, J; Cen, W R; Chan, Y L; Chang, J F; Chang, L C; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J -H; Cheng, J; Cheng, Y P; Cheng, Z K; Cherwinka, J J; Chu, M C; Chukanov, A; Cummings, J P; de Arcos, J; Deng, Z Y; Ding, X F; Ding, Y Y; Diwan, M V; Dolgareva, M; Dove, J; Dwyer, D A; Edwards, W R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guan, M Y; Guo, L; Guo, R P; Guo, X H; Guo, Z; Hackenburg, R W; Han, R; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hor, Y K; Hsiung, Y B; Hu, B Z; Hu, T; Hu, W; Huang, E C; Huang, H X; Huang, X T; Huber, P; Huo, W; Hussain, G; Jaffe, D E; Jaffke, P; Jen, K L; Jetter, S; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Joshi, J; Kang, L; Kettell, S H; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Kwok, T; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lee, J H C; Lei, R T; Leitner, R; Li, C; Li, D J; Li, F; Li, G S; Li, Q J; Li, S; Li, S C; Li, W D; Li, X N; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, S; Lin, S K; Lin, Y -C; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, D W; Liu, J L; Liu, J C; Loh, C W; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Lv, Z; Ma, Q M; Ma, X Y; Ma, X B; Ma, Y Q; Malyshkin, Y; Caicedo, D A Martinez; McDonald, K T; McKeown, R D; Mitchell, I; Mooney, M; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ning, Z; Ochoa-Ricoux, J P; Olshevskiy, A; Pan, H -R; Park, J; Patton, S; Pec, V; Peng, J C; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Raper, N; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Steiner, H; Sun, G X; Sun, J L; Tang, W; Taychenachev, D; Treskov, K; Tsang, K V; Tull, C E; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, C -H; Wu, Q; Wu, W J; Xia, D M; Xia, J K; Xing, Z Z; Xu, J Y; Xu, J L; Xu, Y; Xue, T; Yang, C G; Yang, H; Yang, L; Yang, M S; Yang, M T; Ye, M; Ye, Z; Yeh, M; Young, B L; Yu, Z Y; Zeng, S; Zhan, L; Zhang, C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, X T; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhao, Q W; Zhao, Y B; Zhong, W L; Zhou, L; Zhou, N; Zhuang, H L; Zou, J H

    2016-01-01

    A new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GW$_{\\mathrm{th}}$ nuclear reactors and detected by eight antineutrino detectors deployed in two near (510~m and 560~m flux-weighted baselines) and one far (1580~m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be $0.946\\pm0.020$ ($0.992\\pm0.021$) for the Huber+Mueller (ILL+Vogel) model. A 2.9 $\\sigma$ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4-6~MeV was found in the measured spectrum, with a local significance of 4.4 $\\sigma$. A reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent p...

  6. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

    Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

    2001-07-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  7. Final report of the HFIR (High Flux Isotope Reactor) irradiation facilities improvement project

    Energy Technology Data Exchange (ETDEWEB)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987.

  8. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  9. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    OpenAIRE

    Shirvan, Koroush; Kazimi, Mujid

    2016-01-01

    A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and ...

  10. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation.

  11. Estimating renewable water flux from landscape features

    Science.gov (United States)

    Peterson, Heidi; Nieber, John; Kanivetsky, Roman; Shmagin, Boris

    2010-05-01

    hydrologic response. Using this system-based approach, identifying the watersheds within each regime and using the characteristic data for that watershed set, the structure of connections can be determined and the quantitative influence that terrestrial landscape characteristics have on the renewable flux of the system could be established. Since water balance characteristics vary spatially and temporally, applying this regionalization method to first identify hydrologic regimes and then identify the response associated with specific characteristics, the influence of environmental changes could also be estimated for un-gauged watersheds. Quantifying the spatial impact of environmental change on the natural flux of the system is critical for moving from proverbial "protection" or "environmental improvement" to sustainability of water resources.

  12. Materials for high performance light water reactors

    Science.gov (United States)

    Ehrlich, K.; Konys, J.; Heikinheimo, L.

    2004-05-01

    A state-of-the-art study was performed to investigate the operational conditions for in-core and out-of-core materials in a high performance light water reactor (HPLWR) and to evaluate the potential of existing structural materials for application in fuel elements, core structures and out-of-core components. In the conventional parts of a HPLWR-plant the approved materials of supercritical fossil power plants (SCFPP) can be used for given temperatures (⩽600 °C) and pressures (≈250 bar). These are either commercial ferritic/martensitic or austenitic stainless steels. Taking the conditions of existing light water reactors (LWR) into account an assessment of potential cladding materials was made, based on existing creep-rupture data, an extensive analysis of the corrosion in conventional steam power plants and available information on material behaviour under irradiation. As a major result it is shown that for an assumed maximum temperature of 650 °C not only Ni-alloys, but also austenitic stainless steels can be used as cladding materials.

  13. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  14. Environmentally assisted cracking in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  15. NUCLEAR SUPERHEATER FOR BOILING WATER REACTOR

    Science.gov (United States)

    Holl, R.J.; Klecker, R.W.; Graham, C.B.

    1962-05-15

    A description is given of a boiling water reactor having a superheating region integral with the core. The core consists essentially of an annular boiling region surrounding an inner superheating region. Both regions contain fuel elements and are separated by a cylindrical wall, perforations being provided in the lower portion of the cylindrical wall to permit circulation of a common water moderator between the two regions. The superheater region comprises a plurality of tubular fuel assemblies through which the steam emanating from the boiling region passes to the steam outlet. Each superheater fuel assembly has an outer double-walled cylinder, the double walls being concentrically spaced and connected together at their upper ends but open at the bottom to provide for differential thermal expansion of the inner and outer walls. Gas is entrapped in the annulus between the walls which acts as an insulating space between the fissionable material inside and the moderator outside. (AEC)

  16. Undermoderated spectrum MOX core study. Supercritical pressure light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki; Kosizuka, Seiichi [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    1998-09-01

    The supercritical pressure light water cooling reactor is a nuclear reactor concept with the once through type and the direct cycle reactor cooled with supercritical pressure water. The cooling water controlled with the feed pump flows directly to the turbine and a recirculation is never done by the nuclear reactor of this type. Therefore, this system isn`t equipped with the recirculation system and the steam separator, the system becomes simple. As for this system, it is expected that the cost performance improves. Here, the outline of former study is described. (author)

  17. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  18. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  19. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

    This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

  20. Strategic Targeting of Light Water Reactors.

    Science.gov (United States)

    1982-03-01

    Effects of Nuclear Weapons. U.S. Department of Defense and U.S. Department of Energy, 1977. 15. Glasstone, Samuel and Walter H. Jordan . Nuclear Power and...0 ’Nhi Mi V ni M 10 NN -4 M2 0 M 0-42(N N Mi 𔃺 - i hi V 10M N (N M2𔃾 - 0 0 0’ 1 0 %r N -T0 m - q- 0 It 14 qTm 4T mi hi m’ 10m N m 02 m 0- m 0 m -i...AFIT/GNE/PF/8ZM- 7 4. TITLE (and Subtitle) S.’ T’ PE OF REPORT & PERIOD COVERED STRATEGIC TARGETING MS THESIS OF LIGUfl WATER REACTORS 6. PERFORMING OIG

  1. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor.

    Science.gov (United States)

    Yavar, A R; Sarmani, S B; Wood, A K; Fadzil, S M; Radir, M H; Khoo, K S

    2011-05-01

    Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor. Copyright © 2011 Elsevier Ltd. All rights reserved.

  2. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  3. Improved measurement of the reactor antineutrino flux and spectrum at Daya Bay

    Science.gov (United States)

    An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Cen, W. R.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.-H.; Cheng, J.; Cheng, Y. P.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guan, M. Y.; Guo, L.; Guo, R. P.; Guo, X. H.; Guo, Z.; Hackenburg, R. W.; Han, R.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Joshi, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Lv, Z.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Mooney, M.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. Y.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. B.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

    2017-01-01

    A new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GWth nuclear reactors and detected by eight antineutrino detectors deployed in two near (560 m and 600 m flux-weighted baselines) and one far (1640 m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be 0.946±0.020 (0.992±0.021) for the Huber+Mueller (ILL+Vogel) model. A 2.9σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4-6 MeV was found in the measured spectrum, with a local significance of 4.4σ. A reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent predictions. Supported in part by the Ministry of Science and Technology of China, the United States Department of Energy, the Chinese Academy of Sciences, the CAS Center for Excellence in Particle Physics, the National Natural Science Foundation of China, the Guangdong provincial government, the Shenzhen municipal government, the China General Nuclear Power Group, the Research Grants Council of the Hong Kong Special Administrative Region of China, the MOST and MOE in Taiwan, the U.S. National Science Foundation, the Ministry of Education, Youth and Sports of the Czech Republic, the Joint Institute of Nuclear Research in Dubna, Russia, the NSFC-RFBR joint research program, the National Commission for Scientific and Technological Research of Chile

  4. Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

  5. Measurement of the Reactor Antineutrino Flux and Spectrum at Daya Bay

    CERN Document Server

    An, F P; Band, H R; Bishai, M; Blyth, S; Butorov, I; Cao, D; Cao, G F; Cao, J; Cen, W R; Chan, Y L; Chang, J F; Chang, L C; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J H; Cheng, J; Cheng, Y P; Cherwinka, J J; Chu, M C; Cummings, J P; de Arcos, J; Deng, Z Y; Ding, X F; Ding, Y Y; Diwan, M V; Dove, J; Draeger, E; Dwyer, D A; Edwards, W R; Ely, S R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guan, M Y; Guo, L; Guo, X H; Hackenburg, R W; Han, R; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hor, Y K; Hsiung, Y B; Hu, B Z; Hu, L M; Hu, L J; Hu, T; Hu, W; Huang, E C; Huang, H X; Huang, X T; Huber, P; Hussain, G; Jaffe, D E; Jaffke, P; Jen, K L; Jetter, S; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Kang, L; Kettell, S H; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Kwok, T; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lei, R T; Leitner, R; Leung, K Y; Leung, J K C; Lewis, C A; Li, D J; Li, F; Li, G S; Li, Q J; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, P Y; Lin, S K; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, D W; Liu, H; Liu, J L; Liu, J C; Liu, S S; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, Q M; Ma, X Y; Ma, X B; Ma, Y Q; Caicedo, D A Martinez; McDonald, K T; McKeown, R D; Meng, Y; Mitchell, I; Kebwaro, J Monari; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ning, Z; Ochoa-Ricoux, J P; Olshevski, A; Pan, H -R; Park, J; Patton, S; Pec, V; Peng, J C; Piilonen, L E; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Raper, N; Ren, B; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Shao, B B; Steiner, H; Sun, G X; Sun, J L; Tang, W; Taychenachev, D; Tsang, K V; Tull, C E; Tung, Y C; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, W W; Wang, X; Wang, Y F; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, Q; Xia, D M; Xia, J K; Xia, X; Xing, Z Z; Xu, J Y; Xu, J L; Xu, J; Xu, Y; Xue, T; Yan, J; Yang, C G; Yang, L; Yang, M S; Yang, M T; Ye, M; Yeh, M; Young, B L; Yu, G Y; Yu, Z Y; Zang, S L; Zhan, L; Zhang, C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, Y M; Zhang, Y X; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhao, Q W; Zhao, Y F; Zhao, Y B; Zheng, L; Zhong, W L; Zhou, L; Zhou, N; Zhuang, H L; Zou, J H

    2015-01-01

    This Letter reports a measurement of the flux and energy spectrum of electron antineutrinos from six 2.9~GW$_{th}$ nuclear reactors with six detectors deployed in two near (effective baselines 512~m and 561~m) and one far (1,579~m) underground experimental halls in the Daya Bay experiment. Using 217 days of data, 296,721 and 41,589 inverse beta decay (IBD) candidates were detected in the near and far halls, respectively. The measured IBD yield is (1.55 $\\pm$ 0.04) $\\times$ 10$^{-18}$~cm$^2$/GW/day or (5.92 $\\pm$ 0.14) $\\times$ 10$^{-43}$~cm$^2$/fission. This flux measurement is consistent with previous short-baseline reactor antineutrino experiments and is $0.946\\pm0.022$ ($0.991\\pm0.023$) relative to the flux predicted with the Huber+Mueller (ILL+Vogel) fissile antineutrino model. The measured IBD positron energy spectrum deviates from both spectral predictions by more than 2$\\sigma$ over the full energy range with a local significance of up to $\\sim$4$\\sigma$ between 4-6 MeV. A reactor antineutrino spectrum...

  6. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  7. Water inventory management in condenser pool of boiling water reactor

    Science.gov (United States)

    Gluntz, Douglas M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  8. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P.C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  9. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

    Energy Technology Data Exchange (ETDEWEB)

    Hellfeld, D. [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Bernstein, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marianno, C. [Texas A & M Univ., College Station, TX (United States). Dept. of Nuclear Engineering

    2017-01-01

    The potential of elastic antineutrino-electron scattering (ν¯e + e → ν¯e + e) in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor. Background was estimated via independent simulations and by appropriately scaling published measurements from similar detectors. Many potential backgrounds were considered, including solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclide and water-borne radon decays, and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. The detector response was modeled using a GEANT4-based simulation package. The results indicate that with the use of low radioactivity PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. The directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of theoretical conditions that, if satisfied in practice, would enable nuclear reactor antineutrino directionality in a Gd-doped water Cherenkov detector approximately 10 km from a large power reactor.

  10. Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization.

    Science.gov (United States)

    Snoj, L; Trkov, A; Jaćimović, R; Rogan, P; Zerovnik, G; Ravnik, M

    2011-01-01

    In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)-Gold (0.1 wt%) foils (disks of 5mm diameter and 0.2mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization. Copyright © 2010 Elsevier Ltd. All rights reserved.

  11. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

    Science.gov (United States)

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

    2015-01-01

    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. Copyright © 2014. Published by Elsevier B.V.

  12. A simple model of reactor cores for reactor neutrino flux calculations for the KamLAND experiment

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan)]. E-mail: kyo@awa.tohoku.ac.jp; Inoue, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Owada, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Suekane, F. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Suzuki, A. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Hirano, G. [TEPCO Systems Corporation, Tokyo 135-0034 (Japan); Kosaka, S. [TEPCO Systems Corporation, Tokyo 135-0034 (Japan); Ohta, T. [Tokyo Electric Power Company, Tokyo 100-8560 (Japan); Tanaka, H. [Tokyo Electric Power Company, Tokyo 100-8560 (Japan)

    2006-12-21

    KamLAND is a reactor neutrino oscillation experiment with a very long baseline. This experiment successfully measured oscillation phenomena of reactor antineutrinos coming mainly from 53 reactors in Japan. In order to extract the results, it is necessary to accurately calculate time-dependent antineutrino spectra from all the reactors. A simple model of reactor cores and code implementing it were developed for this purpose. This paper describes the model of the reactor cores used in the KamLAND reactor analysis.

  13. Environmentally assisted cracking in light water reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2007-11-06

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature

  14. Localized fast neutron flux enhancement for damage experiments in a research reactor; Accroissement local du flux rapide pour des experiences de dommages dans un reacteur de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, F

    2003-06-01

    In irradiation experiments on materials in the core of the Osiris reactor (CEA-Saclay) we seek to increase damage in irradiated samples and to reduce the duration of their stay in the core. Damage is essentially caused by fast neutrons (E {>=} 1 MeV); we have therefore pursued the possibility of a localized increase of their level in an irradiation experiment by using a flux converter device made up of fissile material arranged according to a suitable geometry that allows the converter to receive experiments. We have studied several parameters that are influential in the increase of fast neutron flux within the converter. We have also considered the problem of the converter's cooling in the core and its effect on the operation of the reactor. We have carried out a specific neutron calculation scheme based on the modular 2D-transport code APOLLO2 using a two-level transport method. Experimental validation of the flux calculation scheme was carried out in the ISIS reactor, the mock-up of OSIRIS, by optimizing the loading of fuel elements in the core. The experimental results show that the neutron calculation scheme computes the fluxes in close agreement with the measurements especially the fast flux. This study allows us to master the essential physical parameters needed for the design of a flux converter in an MTR reactor. (author)

  15. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  16. A Two‐Fluid model study of hydrogen production via water gas shift in fluidized bed membrane reactors

    OpenAIRE

    J.W. Voncken, Ramon; Roghair, Ivo; Van Sint Annaland, Martin

    2017-01-01

    Fluidized bed membrane reactors have been proposed as a promising reactor concept for the production of ultra-pure hydrogen via Water Gas Shift (WGS). High-flux thin-film dense palladium-based membranes are used to selectively extract hydrogen from the reaction medium, which shifts the thermodynamic equilibrium towards the products’ side, increasing the conversion. A Two-Fluid Model (TFM) has been used to investigate the effect of hydrogen extraction via perm-selective membranes on the WGS re...

  17. Prediction of Reactor Vessel Water Level Using Fuzzy Neural Networks in Severe Accidents due to LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soonho; Kim, Jaehawn; Na, Mangyun [Chosun Univ., Gwangju (Korea, Republic of)

    2013-05-15

    When the initial events that may lead to the severe accident such as Loss Of Coolant Accident (LOCA) and Steam Generator Tube Rupture (SGTR) occurs at a nuclear power plant, it is most important to check the status of the plant conditions by observing the safety-related parameters such as neutron flux, pressurizer pressure, steam generator pressure and water level. In this paper, we propose a method of predicting the water level of coolant in the reactor vessel that directly affect the important events such as the exposure of the reactor core and the damage of reactor vessel by using a Fuzzy Neural Network (FNN) method. In addition, the data for verifying a proposed model was obtained by simulating the severe accident scenarios for the OPR1000 nuclear power plant using the MAAP4 code. In this paper, a prediction model was developed for predicting the reactor vessel water level using the FNN method. The proposed FNN model was verified based on the simulation data of OPR1000 by using MAAP4 code. As a result of simulation, we could see that the performance of the proposed FNN model is quite satisfactory but some large errors are observed occasionally. If the proposed FNN model is optimized by using a variety of data, it is possible to predict the reactor vessel water level exactly.

  18. Verification of Monte Carlo calculations of the neutron flux in typical irradiation channels of the TRIGA reactor, Ljubljana

    NARCIS (Netherlands)

    Jacimovic, R; Maucec, M; Trkov, A

    2003-01-01

    An experimental verification of Monte Carlo neutron flux calculations in typical irradiation channels in the TRIGA Mark II reactor at the Jozef Stefan Institute is presented. It was found that the flux, as well as its spectral characteristics, depends rather strongly on the position of the

  19. Coil Designs for Novel Magnetic Geometries to Cure the Divertor Heat Flux Problem for Reactors

    Science.gov (United States)

    Pekker, M.; Valanju, P.; Kotschenreuther, M.; Wiley, J. C.; Strickler, D.

    2004-11-01

    Coil designs are developed for novel magnetic divertor geometries with a second axi-symmetric x-point and flux expansion region along the separatrix. Adjacent posters describe how these lead to spreading of heat flux and the possibility of stable, complete detachment to overcome serious physics and engineering problems in reactors. The principal feasibility issue is creating, with simple coils, additional X-points on the separatrix without extensively deforming the magnetic field in the main plasma. For the spherical tokamak NSTX, we show that adding one or two poloidal coils suffices to create a divergent flux at the divertor, i.e., a new x-point. The currents and forces for the extra coils are small. We also modify ARIES ST design to show reactor feasibility. Optimized coil designs for PEGASUS, ARIES RS/AT, and a modular ITER retrofit are also being developed. For our calculations we used self consistent code FBEQ, which was used to design NSTX. We also use NCSX tools for optimization of designs with competing physics and engineering constraints.

  20. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.; RECINIELLO,R.N.; HU,J.P.; RORER,D.C.

    2002-08-18

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex{trademark} polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup.

  1. Measurement of neutron flux at the Brazilian IPEN-MB-01 reactor; Medida do fluxo de neutrons no reator IPEN MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Tolentino, Fabiola A. [Universidade Federal do ABC (UFABC) Santo Andre, SP (Brazil); Maiorino, Jose R., E-mail: joserubens.maiorino@ufabc.edu.br [Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas, Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil); D' Bitelli, Ulysses, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This paper approaches the application of the neutron foils technique for measurement of thermal and intermediary neutron fluxes activation in a flux trap placed at the center of reactor core of the IPEN MB-01 reactor. The measurements were made irradiating the naked foils of {sup 197}Au and covered with cadmium foils at selected positions of the water trap flux. The radiation counting induced in the reaction {sup 197}Au(n,{gamma}){sup 198}Au was made by gamma e spectrometry using NaI and HPGe detectors. For obtaining the thermal and intermediary neutron flux the cadmium ratio technique was used. The calibration of gamma e spectrometers was made by using a standard {sup 152}Eu source. Correction for the efficiency of detectors were taking into consideration. The results for the axial and radial flux distributions obtained by the spectrometers are exhibited giving good concordance in the results. As main conclusion, the work allowed the apprenticeship of experimental techniques of measurement of thermal and intermediary fluxes, and the necessary gamma e spectrometry for such measurements. (author)

  2. Capital Cost: Pressurized Water Reactor Plant Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate.

  3. Process for treating effluent from a supercritical water oxidation reactor

    Science.gov (United States)

    Barnes, Charles M.; Shapiro, Carolyn

    1997-01-01

    A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor.

  4. Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

    Science.gov (United States)

    An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Chan, Y. L.; Chang, J. F.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guo, L.; Guo, X. H.; Guo, Y. H.; Guo, Z.; Hackenburg, R. W.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huang, Y. B.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jen, K. L.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Kang, L.; Kettell, S. H.; Khan, A.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Qiu, R. M.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Stoler, P.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Yang, Y. Z.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, C. C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, R.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhou, L.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

    2017-06-01

    The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 G Wth reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F239 from 0.25 to 0.35, Daya Bay measures an average IBD yield σ¯f of (5.90 ±0.13 )×10-43 cm2/fission and a fuel-dependent variation in the IBD yield, d σf/d F239, of (-1.86 ±0.18 )×10-43 cm2/fission . This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1 σ . This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17 ±0.17 ) and (4.27 ±0.26 )×10-43 cm2 /fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

  5. Development of a Scale Model for High Flux Isotope Reactor Cycle 400

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Dan [ORNL

    2012-03-01

    The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

  6. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  7. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

    Energy Technology Data Exchange (ETDEWEB)

    Hellfeld, D., E-mail: dhellfeld@berkeley.edu [Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA 94720 (United States); Bernstein, A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Dazeley, S., E-mail: dazeley2@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Marianno, C. [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77843 (United States)

    2017-01-01

    The potential of elastic antineutrino-electron scattering in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13-km standoff from a 3.758-GWt light water nuclear reactor and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays from the photomultiplier tubes (PMTs), detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. The results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3σ significance in large Gd-doped water Cherenkov detectors with greater than 10-km standoff from a nuclear reactor.

  8. Core design analysis of the supercritical water fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mori, M.

    2005-10-01

    Light Water Reactor technology is nowadays the most successful commercial application of fission reactors for the production of electricity. However, in the next years, nuclear industry will have to face new and demanding challenges. The need for sustainable and cheap sources of energy, the need for public acceptance, the need for even higher safety standards, the need to minimize waste production are only a few examples. It is for these very reasons that a few next generation nuclear reactor concepts were selected for extensive research and development. Super critical water cooled reactors are one of them. The use of a supercritical coolant would in fact allow for higher thermal efficiencies and a more compact plant design. As a matter of fact, steam generators, or steam separators and driers would not be needed thus, significantly reducing construction costs. Moreover, because of the high heat capacity of supercritical water, comparatively less coolant would be needed to refrigerate the reactor. Consequently, a water-cooled reactor with a fast neutron spectrum could potentially be designed: the SuperCritical water Fast Reactor. This system presents unique features combining well-known fast and light water reactor characteristics in one design (e.g. the tendency to a positive void reactivity coefficient together with Loss Of Coolant Accidents, as design basis). The core is in fact loaded with highly enriched Mixed Oxide fuel (average plutonium content of {approx}23%), and presents a peculiar and significant geometrical and material heterogeneity (use of radial and axial blankets, solid moderator layers, several enrichment zones). The safety analysis of this very complex core layout, the development of suitable tools of investigation, and the optimization of the void reactivity effect through core design, is the main objective of this work. (orig.)

  9. Secondary flows in the cooling channels of the high-performance light-water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Laurien, E.; Wintterle, Th. [Stuttgart Univ., Institute for Nuclear Technolgy and Energy Systems (IKE) (Germany)

    2007-07-01

    The new design of a High-Performance Light-Water Reactor (HPLWR) involves a three-pass core with an evaporator region, where the compressed water is heated above the pseudo-critical temperature, and two superheater regions. Due to the strong dependency of the supercritical water density on the temperature significant mass transfer between neighboring cooling channels is expected if the temperature is unevenly distributed across the fuel element. An inter-channel flow is then superimposed to the secondary flow vortices induced by the non-isotropy of turbulence. In order to gain insight into the resulting flow patterns as well as into temperature and density distributions within the various subchannels of the fuel element CFD (Computational Fluid Dynamics) calculations for the 1/8 fuel element are performed. For simplicity adiabatic boundary conditions at the moderator box and the fuel element box are assumed. Our investigation confirms earlier results obtained by subchannel analysis that the axial mass flux is significantly reduced in the corner subchannel of this fuel element resulting in a net mass flux towards the neighboring subchannels. Our results provide a first estimation of the magnitude of the secondary flows in the pseudo-critical region of a supercritical light-water reactor. Furthermore, it is demonstrated that CFD is an efficient tool for investigations of flow patterns within nuclear reactor fuel elements. (authors)

  10. Quarterly technical report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, July--September 1975

    Energy Technology Data Exchange (ETDEWEB)

    1976-02-01

    Light water reactor safety activities performed during July through September 1975 are summarized. The isothermal blowdown test series of the Semiscale Mod-1 test program has provided data for evaluation of break flow phenomena and analyses of piping flow regimes and pump performance. In the LOFT Program, measurement uncertainties were evaluated. The Thermal Fuels Behavior Program completed two power-cooling-mismatch tests on PWR-type fuel rods to investigate critical heat flux characteristics. Model development and verification efforts of the Reactor Behavior Program included development of the SPLEN1 computer code, subroutines for the FRAP-T code, verification of RELAP4, and results of the Halden Recycle Plutonium Experiment.

  11. Neutron flux characterisation of the Pavia TRIGA Mark II research reactor for radiobiological and microdosimetric applications.

    Science.gov (United States)

    Alloni, D; Prata, M; Salvini, A; Ottolenghi, A

    2015-09-01

    Nowadays the Pavia TRIGA reactor is available for national and international collaboration in various research fields. The TRIGA Mark II nuclear research reactor of the Pavia University offers different in- and out-core neutron irradiation channels, each characterised by different neutron spectra. In the last two years a campaign of measurements and simulations has been performed in order to guarantee a better characterisation of these different fluxes and to meet the demands of irradiations that require precise information on these spectra in particular for radiobiological and microdosimetric studies. Experimental data on neutron fluxes have been collected analysing and measuring the gamma activity induced in thin target foils of different materials irradiated in different TRIGA experimental channels. The data on the induced gamma activities have been processed with the SAND II deconvolution code and finally compared with the spectra obtained with Monte Carlo simulations. The comparison between simulated and measured spectra showed a good agreement allowing a more precise characterisation of the neutron spectra and a validation of the adopted method. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  12. Light Water Reactor Sustainability Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  13. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  14. Assessment of the high performance light water reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J. [Univ. of Stuttgart, IKE, (Germany); Schulenberg, T. [Karlsruhe Inst. of Tech., Karlsruhe (Germany); Bittermann, D. [AREVA NP GmbH, Erlangen (Germany); Andreani, M. [Paul Scherrer Inst., Villigen (Switzerland); Maraczy, C. [AEKI-KFKI, Budapest (Hungary)

    2011-07-01

    From 2006-2010, the High Performance Light Water Reactor (HPLWR) was investigated within a European Funded project called HPLWR Phase 2. Operated at 25MPa with a heat-up rate in the core from 280{sup o}C to 500{sup o}C, this reactor concept provides a technological challenge in the fields of design, neutronics, thermal-hydraulics and heat transfer, materials, and safety. The assessment of the concept with respect to the goals of the technology roadmap for Generation IV Nuclear Reactors of the Generation IV International Forum shows that the HPLWR has a potential to fulfil the goals of economics, safety and proliferation resistance and physical protection. In terms of sustainability, the HPLWR with a thermal neutron spectrum investigated within this project, does not differ from existing Light Water Reactors in terms of usage of fuel and waste production. (author)

  15. Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

    1981-09-01

    The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.

  16. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  17. Studies of low temperature, low flux radiation embrittlement of nuclear reactor structural materials. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Lucas, G.E.

    1998-09-02

    A large matrix of simple alloys and complex commercial type steels was irradiated over a range of fluxes at 60 C up to a fast fluence of about 3 {times} 10{sup 22} n/m{sup 2}. Combined with data in the literature, these results show a negligible effect of flux on irradiation hardening in the range of 2 {times} 10{sup 13} to 5 {times} 10{sup 18} n/m{sup 2}-s. This observation lends indirect support to the proposal that the accelerated embrittlement in the High Flux Isotope Reactor surveillance steels was due to an anomalously high level of damage from gamma rays. A weak dependence of hardening on a number of elements, including copper, nickel, phosphorus, molybdenum and manganese, can be described by a simple empirical chemistry factor. Particular combinations of elements resulted in hardening differences of up to about 60% in the complex commercial type steels and up to about 100% in simple model alloys. Direct effects of microstructure appear to be minimal. Hardening varies with the square root of fluence above a threshold around 4 {times} 10{sup 20} n/m{sup 2}. The results suggest that low temperature hardening is dominated by local intracascade processes leading to the formation of small defect-solute clusters/complexes. The observed hardening corresponds to nominal maximum end-of-life transition temperature shifts in support structure steels of about 120 C.

  18. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    Energy Technology Data Exchange (ETDEWEB)

    Austin, W.; Brinkley, D.

    2010-05-05

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these

  19. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  20. Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield.

    Science.gov (United States)

    Merz, Stefan; Djuricic, Mile; Villa, Mario; Böck, Helmuth; Steinhauser, Georg

    2011-11-01

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10(9)cm(-2)s(-1) at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. Copyright © 2011 Elsevier Ltd. All rights reserved.

  1. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  2. 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.

  3. Design of virtual SCADA simulation system for pressurized water reactor

    Science.gov (United States)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  4. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  5. Novel Photocatalytic Reactor Development for Removal of Hydrocarbons from Water

    Directory of Open Access Journals (Sweden)

    Morgan Adams

    2008-01-01

    Full Text Available Hydrocarbons contamination of the marine environment generated by the offshore oil and gas industry is generated from a number of sources including oil contaminated drill cuttings and produced waters. The removal of hydrocarbons from both these sources is one of the most significant challenges facing this sector as it moves towards zero emissions. The application of a number of techniques which have been used to successfully destroy hydrocarbons in produced water and waste water effluents has previously been reported. This paper reports the application of semiconductor photocatalysis as a final polishing step for the removal of hydrocarbons from two waste effluent sources. Two reactor concepts were considered: a simple flat plate immobilised film unit, and a new rotating drum photocatalytic reactor. Both units proved to be effective in removing residual hydrocarbons from the effluent with the drum reactor reducing the hydrocarbon content by 90% under 10 minutes.

  6. Selection of support structure materials for irradiation experiments in the HFIR (High Flux Isotope Reactor) at temperatures up to 500 degrees C

    Energy Technology Data Exchange (ETDEWEB)

    Farrell, K.; Longest, A.W.

    1990-01-01

    The key factor in the design of capsules for irradiation of test specimens in the High Flux Isotope Reactor at preselected temperatures up to 500{degree}C utilizing nuclear heating is a narrow gas-filled gap which surrounds the specimens and controls the transfer of heat from the specimens through the wall of a containment tube to the reactor cooling water. Maintenance of this gap to close tolerances is dependent on the characteristics of the materials used to support the specimens and isolate them from the water. These support structure materials must have low nuclear heating rates, high thermal conductivities, and good dimensional stabilities under irradiation. These conditions are satisfied by certain aluminum alloys. One of these alloys, a powder metallurgy product containing a fine dispersion of aluminum oxide, is no longer manufactured. A new alloys of this type, with the trade name DISPAL, is determined to be a suitable substitute. 23 refs., 13 figs., 3 tabs.

  7. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  8. Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

    Science.gov (United States)

    Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

    2012-10-01

    A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

  9. REFLECTOR CONTROL OF A BOILING-WATER REACTOR

    Science.gov (United States)

    Treshow, M.

    1962-05-22

    A line connecting the reactor with a spent steam condenser contains a valve set to open when the pressure in the reactor exceeds a predetermined value and an orifice on the upstream side of the valve. Another line connects the reflector with this line between the orifice and the valve. An excess steam pressure causes the valve to open and the flow of steam through the line draws water out of the reflector. Provision is also made for adding water to the reflector when the steam pressure drops. (AEC)

  10. Transmutation of plasma facing materials by the neutron flux in a DT fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Behrisch, R. [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Khripunov, V.; Santoro, R.T. [The ITER JCT, Max Planck Inst. fuer Plasmaphysik, Garching (Germany); Yesil, J.M

    1998-10-01

    The neutron induced transmutations, gas production, activation, atomic displacements and afterheat have been calculated for several potential low- and high-Z first wall materials in a DT fusion reactor. The neutron spectra taken are those calculated for the first wall of the shielding blanket proposed for the international thermonuclear experimental reactor (ITER). For low Z materials the production of hydrogen and helium isotopes dominate, while for high Z materials activation and afterheat dominate. For a total flux of 14.1 MeV neutrons corresponding to a wall load of 1 MW a m{sup -2}, the tritium production in a 1 cm thick low Z material layer of the first wall is of the order of 10{sup 16} - 10{sup 19} cm{sup -2}, while the calculated He concentrations are of the order of {proportional_to}10{sup -3} per wall atom or 10{sup 20} cm{sup -2}. These gases may modify the thermophysical properties of the wall material. (orig.) 34 refs.

  11. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events refs., 139 tabs., 85 figs. Prepared for Department of Industry, Science and Tourism

  12. Self-Sustaining Thorium Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States); Gorman, Phillip M. [Univ. of California, Berkeley, CA (United States); Bogetic, Sandra [Univ. of California, Berkeley, CA (United States); Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States); Zhang, Guanheng [Univ. of California, Berkeley, CA (United States); Varela, Christopher R. [Univ. of California, Berkeley, CA (United States); Fratoni, Massimiliano [Univ. of California, Berkeley, CA (United States); Vijic, Jasmina J. [Univ. of California, Berkeley, CA (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Hall, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Ward, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Jarrett, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Wysocki, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Xu, Yunlin [Univ. of Michigan, Ann Arbor, MI (United States); Kazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Shirvan, Koroush [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mieloszyk, Alexander [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todosow, Michael [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, Nicolas [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  13. Ingested water equilibrates isotopically with the body water pool of a shorebird with unrivaled water fluxes

    NARCIS (Netherlands)

    Visser, G.H.; Dekinga, A; Achterkamp, B.; Piersma, T.

    2000-01-01

    We investigated the applicability of H-2 to measure the amount of body water (TBW) and water fluxes in relation to diet type and level of food intake in a mollusk-eating shorebird, the Red Knot (Calidris canutus). Six birds were exposed to eight experimental indoor conditions. Average fractional H-2

  14. Job/task analysis for I C (Instrumentation and Controls) instrument technicians at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duke, L.L.

    1989-09-01

    To comply with Department of Energy Order 5480.XX (Draft), a job/task analysis was initiated by the Maintenance Management Department at Oak Ridge National Laboratory (ORNL). The analysis was applicable to instrument technicians working at the ORNL High Flux Isotope Reactor (HFIR). This document presents the procedures and results of that analysis. 2 refs., 2 figs.

  15. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  16. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  17. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    Energy Technology Data Exchange (ETDEWEB)

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-07-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods.

  18. Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Betzler, Ben [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Hirtz, Gregory John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Sunny, Eva [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-09-01

    The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se production capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

  19. Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

    Directory of Open Access Journals (Sweden)

    Noble Brooklyn

    2012-01-01

    Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

  20. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  1. Fuel assembly design study for a reactor with supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Hofmeister, J. [RWE Power AG, Huyssenallee 2, D-45128 Essen (Germany); Waata, C. [ANSYS Germany GmbH, Staudenfeldweg 12, D-83624 Otterfing (Germany); Starflinger, J. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, P.O. Box 3640, D-76021 Karlsruhe (Germany); Schulenberg, T. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, P.O. Box 3640, D-76021 Karlsruhe (Germany)]. E-mail: thomas.schulenberg@iket.fzk.de; Laurien, E. [University of Stuttgart, Institute for Nuclear Technology and Energy Systems (IKE), Pfaffenwaldring 31, D-70569 Stuttgart (Germany)

    2007-08-15

    The European concept of the High Performance Light Water Reactor (HPLWR) differs from current light water reactors in a higher system pressure beyond the critical point of water, as well as a higher heat-up of the coolant within the core and thus higher core outlet temperatures, leading to a significant increase in turbine power and thermal efficiency of the power plant. The motivation to develop a novel fuel assembly for the HPLWR is caused by the high variation of coolant density in the core by more than a factor of seven. A systematic design study shows that a square fuel assembly with two rows of fuel rods and a central moderator box is best to minimize the structural material, to optimize the moderator to fuel ratio and to reduce differences of fuel rod power. Using neutronic and thermal-hydraulic analyses, a detailed mechanical design of a fuel assembly of the HPLWR has been worked out. Moreover, concepts for the head piece, the foot piece, the steam plenum and the lower mixing plenum, including the lower core plate, have been developed to account for the individual flow paths of this reactor. These allow a leak-tight counter current flow of moderator water and coolant as well as uniform mixing of different mass flows. The assembly design concept can be used as a general key component for any advanced core design of this reactor.

  2. Light water reactor fuel response during reactivity initiated accident experiments

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P. E.; McCardell, R. K.; Martinson, Z. R.; Seiffert, S. L.

    1979-01-01

    Experimental results from six recent Power Burst Facility (PBF) reactivity initiated accident (RIA) tests are compared with data from previous Special Power Excursion Reactor Test (SPERT), and Japanese Nuclear Safety Research Reactor (NSRR) tests. The RIA fuel behavior experimental program recently started in the PBF is being conducted with coolant conditions typical of hot-startup conditions in a commercial boiling water reactor. The SPERT and NSRR test programs investigated the behavior of single or small clusters of light water reactor (LWR) type fuel rods under approximate room temperature and atmospheric pressure conditions in capsules containing stagnant water. As observed in the SPERT and NSRR tests, energy deposition, and consequent enthalpy increase in the PBF test fuel, appears to be the single most important variable. However, the consequences of failure at boiling water hot-startup system conditions appear to be more severe than previously observed in either the stagnant capsule SPERT or NSRR tests. Metallographic examination of both previously unirradiated and irradiated PBF fuel rod cross sections revealed extensive variation in cladding wall thicknesses (involving considerable plastic flow) and fuel shattering along grain boundaries in both restructured and unrestructured fuel regions. Oxidation of the cladding resulted in fracture at the location of cladding thinning and disintegration of the rods during quench. In addition,swelling of the gaseous and potentially volatile fission products in previously irradiated fuel resulted in volume increases of up to 180% and blockage of the coolant channels within the flow shrouds surrounding the fuel rods.

  3. Risk management and decision rules for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Griesmeyer, J. M.; Okrent, D.

    1981-01-01

    The process of developing and adopting safety objectives in quantitative terms can provide a basis for focusing societal decision making on the suitability of such objectives and upon questions of compliance with those objectives. A preliminary proposal for a light water reactor (LWR) risk management framework is presented as part of that process.

  4. Plant Control of the High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schlagenhaufer, Marc; Starflinger, J.; Schulenberg, T. [Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe GmbH, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen, Baden-Wuertemberg 76344 (Germany)

    2009-06-15

    The latest design concept of the High Performance Light Water Reactor (HPLWR) includes a thermal core in which supercritical water at 25 MPa inlet pressure is heated up from 280 deg. C reactor inlet temperature to 500 deg. C core exit temperature in three steps with intermediate coolant mixing to minimize peak cladding temperatures of the fuel rods. A direct supercritical steam cycle of the HPLWR has been designed with high, intermediate and low pressure turbines with a single reheat to 441 deg. C at 4.04 MPa pressure. Three low pressure pre-heaters and four high pressure pre-heaters are foreseen to achieve the envisaged reactor inlet temperature of 280 deg. C at full load. A feedwater tank of 603 m{sup 3} at 0.55 MPa pressure serves as an accumulator for normal and accidental conditions. The steam cycle has been modelled with APROS, developed by VTT Finland, to provide thermodynamic data and cycle efficiency values under full load and part load operation conditions as well as the transient response to load changes. A plant control system has been designed in which the reactor inlet pressure is controlled by the turbine valve, the reactor power is controlled by the feedwater pumps while the life steam temperature is controlled by control rods, and the reheat temperature is controlled by the reheater valve. Neglecting the reactivity control, the core power can also be treated as input parameter such that the life steam temperature is directly controlled by the feedwater mass flow. The plant control can handle all loading and de-loading cycles including complete shut down. A constant pressure at reactor inlet is foreseen for all load cases. Peak temperatures of the fuel pins are checked with a simplified core model. Two shut down procedures starting at 50% load are presented. A reactor scram with turbine states the safe shut down of the whole plant. To avoid hard material temperature changes, a controlled shut down procedure is designed. The rotational speed of the

  5. The Consortium for Advanced Simulation of Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  6. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  7. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES...

  8. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR) standard...

  9. Conceptual design of a large heavy water reactor for US siting

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, N L; Jesick, J F

    1979-09-01

    Information is presented concerning fuel management and safety and licensing assessment of the pressurized heavy water reactor; and commercial introduction of the pressurized heavy water reactor in the United States.

  10. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  11. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  12. Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Ade, Brian J [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Sunny, Eva E [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Betzler, Benjamin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Pinkston, Daniel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR)

    2015-03-01

    A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

  13. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

    2011-03-01

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one

  14. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security Requirements 1.0 Background The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by the..., which utilized a forced-circulation, direct-cycle boiling water reactor as its heat source. The plant is...

  15. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  16. The Water Suitcase of Migrants: Assessing Virtual Water Fluxes Associated to Human Migration

    National Research Council Canada - National Science Library

    Metulini, Rodolfo; Tamea, Stefania; Laio, Francesco; Riccaboni, Massimo

    2016-01-01

    .... We show here that such belief is over-simplistic. We reframe the problem by considering the international food trade and the corresponding virtual water fluxes, which quantify the water used for the production of traded agricultural commodities...

  17. Multi-Applications Small Light Water Reactor - NERI Final Report

    Energy Technology Data Exchange (ETDEWEB)

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  18. The Use of Predictive Modeling in the Characterization and Remediation of the High Flux Beam Reactor Tritium Plume

    Science.gov (United States)

    Gunther, W.; Hauptmann, M.; Sullivan, T.

    2008-12-01

    Discovery of tritium in the groundwater near the Brookhaven National Laboratory High Flux Beam Reactor (HFBR) in January 1997 initiated an intense effort to locate its source, determine its extent, and implement remediation methods to mitigate its consequences. The tritium plume that resulted from the slow, long-term seepage of contaminated water from the HFBR spent fuel pool was the subject of much attention by the media, State and Federal regulators, and the public in 1997. Tremendous amounts of resources were dedicated to characterizing the plume in great detail and to mitigating its consequences. An important element of the remediation activities and in communicating the status of these efforts to stakeholders involved the modeling of the plume, particularly its future path and extent of travel. This paper describes the effectiveness of the models that were employed and the assumptions made in predicting the plume migration over time at an active research laboratory site. The prediction that was made by the model for the selected remediation alternative is compared with the actual groundwater monitoring data obtained over 10 years to assess the performance of the model. This paper also describes how the outputs from the groundwater models were used to effectively convey information about the remediation efforts to environmental regulators and the local community.

  19. Design concept of the high performance light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schulenberg, Thomas; Starflinger, Joerg [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. for Nuclear and Energy Technologies; Bittermann, Dietmar [AREVA NP GmbH, Erlangen (Germany). NEP-G Process

    2009-04-15

    The 'High Performance Light Water Reactor' (HPLWR) is a Light Water Reactor operating with supercritical water as coolant. At a pressure of 25 MPa in the core, water is heated up from 280 to 500 C. For these conditions, the envisaged net plant efficiency is 43.5%. The core design concept is based on a so-called '3-pass-core' in which the coolant is heated up in three subsequent steps. After each step, the coolant is mixed avoiding hot streaks possibly leading to unacceptable wall temperatures. The design of such a core comprises fuel assemblies containing 40 fuel rods and an inner and outer box for a better neutron moderation. Nine of these are assembled to a cluster with common head- and foot piece. The coolant is mixed inside an upper and inside a lower mixing chamber and leaves the reactor pressure vessel through a co-axial pipe, which protects the vessel wall against too high temperatures. (orig.)

  20. The HB-2D Polarized Neutron Development Beamline at the High Flux Isotope Reactor

    Science.gov (United States)

    Crow, Lowell; Hamilton, WA; Zhao, JK; Robertson, JL

    2016-09-01

    The Polarized Neutron Development beamline, recently commissioned at the HB-2D position on the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, provides a tool for development and testing of polarizers, polarized neutron devices, and prototyping of polarized neutron techniques. With available monochromators including pyrolytic graphite and polarizing enriched Fe-57 (Si), the instrument has operated at 4.25 and 2.6 Å wavelengths, using crystal, supermirror, or He-3 polarizers and analyzers in various configurations. The Neutron Optics and Development Team has used the beamline for testing of He-3 polarizers for use at other HFIR and Spallation Neutron Source (SNS) instruments, as well as a variety of flipper devices. Recently, we have acquired new supermirror polarizers which have improved the instrument performance. The team and collaborators also have continuing demonstration experiments of spin-echo focusing techniques, and plans to conduct polarized diffraction measurements. The beamline is also used to support a growing use of polarization techniques at present and future instruments at SNS and HFIR.

  1. Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

    Energy Technology Data Exchange (ETDEWEB)

    Magielsen, A.J., E-mail: magielsen@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Jong, M.; Bakker, T.; Luzginova, N.V.; Mutnuru, R.K.; Ketema, D.J.; Fedorov, A.V. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands)

    2011-08-31

    In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 deg. C and 500 deg. C. During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 deg. C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 deg. C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.

  2. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  3. Ion and water fluxes in the ileum of rats.

    Science.gov (United States)

    CURRAN, P F; SOLOMON, A K

    1957-09-20

    Studies have been carried out on the movement of salt and water across the small intestine of the rat. Segments of the ileum of anesthetized rats have been perfused in vivo with unbuffered NaCl solutions or isotonic solutions of NaCl and mannitol. Kinetic analysis of movements of Na(24) and Cl(36) has permitted determination of the efflux and influx of Na and Cl. Net water absorption has been measured using hemoglobin as a reference substance. Water was found to move freely in response to gradients of osmotic pressure. Net water flux from isotonic solutions with varying NaCl concentration was directly dependent on net solute flux. The amount of water absorbed was equivalent to the amount required to maintain the absorbed solute at isotonic concentration. These results have been interpreted as indicating that water movement is a passive process depending on gradients of water activity and on the rate of absorption of solute. The effluxes of Na and Cl are linear functions of concentration in the lumen, but both ions are actively transported by the ileum according to the criterion of Ussing (Acta Physiol. Scand., 1949, 19, 43). The electrical potential difference between the lumen and plasma has been interpreted as a diffusion potential slightly modified by the excess of active Cl flux over active Na flux. The physical properties of the epithelial membrane indicate that it is equivalent to a membrane having negatively charged uniform right circular pores of 36 A radius occupying 0.001 per cent of the surface area.

  4. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  5. Spiral-shaped reactor for water disinfection

    KAUST Repository

    Soukane, Sofiane

    2016-04-20

    Chlorine-based processes are still widely used for water disinfection. The disinfection process for municipal water consumption is usually carried out in large tanks, specifically designed to verify several hydraulic and disinfection criteria. The hydrodynamic behavior of contact tanks of different shapes, each with an approximate total volume of 50,000 m3, was analyzed by solving turbulent momentum transport equations with a computational fluid dynamics code, namely ANSYS fluent. Numerical experiments of a tracer pulse were performed for each design to generate flow through curves and investigate species residence time distribution for different inlet flow rates, ranging from 3 to 12 m3 s−1. A new nature-inspired Conch tank design whose shape follows an Archimedean spiral was then developed. The spiral design is shown to strongly outperform the other tanks’ designs for all the selected plug flow criteria with an enhancement in efficiency, less short circuiting, and an order of magnitude improvement in mixing and dispersion. Moreover, following the intensification philosophy, after 50% reduction in its size, the new design retains its properties and still gives far better results than the classical shapes.

  6. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and/or Mars. These reactors require robust automatic control systems using low mass, rapid...

  7. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and Mars. These reactors require robust automatic control systems using low mass, rapid...

  8. Dense Medium Plasma Water Purification Reactor (DMP WaPR) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Dense Medium Plasma Water Purification Reactor offers significant improvements over existing water purification technologies used in Advanced Life Support...

  9. Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2009-11-01

    An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

  10. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Sease, John D [ORNL; Guida, Tracey [University of Pittsburgh; Jolly, Brian C [ORNL

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  11. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-10-03

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in ancircular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mas velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel.

  12. Fixed-biofilm reactors applied to waste water treatment and aquacultural water recirculating systems

    NARCIS (Netherlands)

    Bovendeur, J.

    1989-01-01

    Fixed-biofilm waste water treatment may be regarded as one of the oldest engineered biological waste water treatment methods. With the recent introduction of modern packing materials, this type of reactor has received a renewed impuls for implementation in a wide field of water treatment.

    In

  13. Advanced Water-Gas Shift Membrane Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sean Emerson; Thomas Vanderspurt; Susanne Opalka; Rakesh Radhakrishnan; Rhonda Willigan

    2009-01-07

    The overall objectives for this project were: (1) to identify a suitable PdCu tri-metallic alloy membrane with high stability and commercially relevant hydrogen permeation in the presence of trace amounts of carbon monoxide and sulfur; and (2) to identify and synthesize a water gas shift catalyst with a high operating life that is sulfur and chlorine tolerant at low concentrations of these impurities. This work successfully achieved the first project objective to identify a suitable PdCu tri-metallic alloy membrane composition, Pd{sub 0.47}Cu{sub 0.52}G5{sub 0.01}, that was selected based on atomistic and thermodynamic modeling alone. The second objective was partially successful in that catalysts were identified and evaluated that can withstand sulfur in high concentrations and at high pressures, but a long operating life was not achieved at the end of the project. From the limited durability testing it appears that the best catalyst, Pt-Re/Ce{sub 0.333}Zr{sub 0.333}E4{sub 0.333}O{sub 2}, is unable to maintain a long operating life at space velocities of 200,000 h{sup -1}. The reasons for the low durability do not appear to be related to the high concentrations of H{sub 2}S, but rather due to the high operating pressure and the influence the pressure has on the WGS reaction at this space velocity.

  14. Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Littrell, Ken [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Peter [Univ. of California, Santa Barbara, CA (United States); Cunningham, Nicholas J. [Univ. of California, Santa Barbara, CA (United States)

    2015-09-01

    The major issues regarding irradiation effects are discussed in [1-3] and have also been discussed in previous progress and milestone reports. As noted previously, of the many significant issues discussed, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. It is clear that embrittlement of RPV steels is a critical issue that may limit LWR plant life extension. The primary objective of the LWRSP RPV task is to develop robust predictions of transition temperature shifts (TTS) at high fluence ( t) to at least 1020 n/cm2 (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. Correlations between the high flux test reactor results and low flux surveillance specimens must be established for proper RPV embrittlement predictions of the current nuclear power fleet. Additionally, a complete understanding of defect evolution for high nickel RPV steels is needed to characterize the embrittlement potential of Mn-Ni-enriched precipitates (MNPs), particularly for the high fluence regime. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the recent discovery and experimental verification [4] of late blooming MNPs with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. New and existing databases will be combined to support developing physically based models of TTS for high fluence-low flux ( < 10 11n/cm2-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1 1020 n/cm2 (>1 MeV). Moreover, large number of various RPV materials have been irradiated in ATR-2 experiment and will be jointly studied by University of California Santa Barbara (UCSB) and ORNL to address majority of microstructural characteristics

  15. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

    The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

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

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

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

  17. Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

    Directory of Open Access Journals (Sweden)

    Shanfang Huang

    2017-01-01

    Full Text Available High thermal neutron fluxes are needed in some research reactors and for irradiation tests of materials. A High Flux Research Reactor (HFRR with an inverse flux trap-converter target structure is being developed by the Reactor Engineering Analysis Lab (REAL at Tsinghua University. This paper studies the safety of the HFRR core by full core flow and temperature calculations using the porous media approach. The thermal nonequilibrium model is used in the porous media energy equation to calculate coolant and fuel assembly temperatures separately. The calculation results show that the coolant temperature keeps increasing along the flow direction, while the fuel temperature increases first and decreases afterwards. As long as the inlet coolant mass flow rate is greater than 450 kg/s, the peak cladding temperatures in the fuel assemblies are lower than the local saturation temperatures and no boiling exists. The flow distribution in the core is homogeneous with a small flow rate variation less than 5% for different assemblies. A large recirculation zone is observed in the outlet region. Moreover, the porous media model is compared with the exact model and found to be much more efficient than a detailed simulation of all the core components.

  18. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR...

  19. Carbon fluxes, evapotranspiration, and water use efficiency of terrestrial ecosystems in China

    Science.gov (United States)

    Jingfeng Xiao; Ge Sun; Jiquan Chen; Hui Chen; Shiping Chen; Gang Dong

    2013-01-01

    The magnitude, spatial patterns, and controlling factors of the carbon and water fluxes of terrestrial ecosystems in China are not well understood due to the lack of ecosystem-level flux observations. We synthesized flux and micrometeorological observations from 22 eddy covariance flux sites across China,and examined the carbon fluxes, evapotranspiration (ET), and...

  20. Reactor process water (PW) piping inspections, 1984--1990

    Energy Technology Data Exchange (ETDEWEB)

    Ehrhart, W.S.; Elder, J.B.; Sprayberry, R.E.; Vande Kamp, R.W.

    1990-12-31

    In July 1983, the NRC ordered the shutdown of five boiling water reactors (BWR`s) because of concerns about reliability of ultrasonic examination for detecting intergranular stress corrosion cracking (IGSCC). These concerns arose because of leaking piping at Niagara Mohawk`s Nine Mile Point which was attributed to IGSCC. The leaks were detected shortly after completion of ultrasonic examinations of the piping. At that time, the Dupont plant manager at Savannah River (SR) directed that investigations be performed to determine if similar problems could exist in SR reactors. Investigation determined that all conditions believed necessary for the initiation and propagation of IGSCC in austenitic stainless steel exist in SR reactor process water (PW) systems. Sensitized, high carbon, austenitic stainless steel, a high purity water system with high levels of dissolved oxygen, and the residual stresses associated with welding during construction combine to provide the necessary conditions. A periodic UT inspection program is now in place to monitor the condition of the reactor PW piping systems. The program is patterned after NRC NUREG 0313, i.e., welds are placed in categories based on their history. Welds in upgraded or replacement piping are examined on a standard schedule (at least every five years) while welds with evidence of IGSCC, evaluated as acceptable for service, are inspected at every extended outage (15 to 18 months). This includes all welds in PW systems three inches in diameter and above. Welds are replaced when MSCC exceeds the replacement criteria of more than twenty percent of pipe circumference of fifty percent of through-wall depth. In the future, we intend to perform flow sizing with automated UT techniques in addition to manual sizing to provide more information for comparison with future examinations.

  1. Reactor process water (PW) piping inspections, 1984--1990

    Energy Technology Data Exchange (ETDEWEB)

    Ehrhart, W.S.; Elder, J.B.; Sprayberry, R.E.; Vande Kamp, R.W.

    1990-01-01

    In July 1983, the NRC ordered the shutdown of five boiling water reactors (BWR's) because of concerns about reliability of ultrasonic examination for detecting intergranular stress corrosion cracking (IGSCC). These concerns arose because of leaking piping at Niagara Mohawk's Nine Mile Point which was attributed to IGSCC. The leaks were detected shortly after completion of ultrasonic examinations of the piping. At that time, the Dupont plant manager at Savannah River (SR) directed that investigations be performed to determine if similar problems could exist in SR reactors. Investigation determined that all conditions believed necessary for the initiation and propagation of IGSCC in austenitic stainless steel exist in SR reactor process water (PW) systems. Sensitized, high carbon, austenitic stainless steel, a high purity water system with high levels of dissolved oxygen, and the residual stresses associated with welding during construction combine to provide the necessary conditions. A periodic UT inspection program is now in place to monitor the condition of the reactor PW piping systems. The program is patterned after NRC NUREG 0313, i.e., welds are placed in categories based on their history. Welds in upgraded or replacement piping are examined on a standard schedule (at least every five years) while welds with evidence of IGSCC, evaluated as acceptable for service, are inspected at every extended outage (15 to 18 months). This includes all welds in PW systems three inches in diameter and above. Welds are replaced when MSCC exceeds the replacement criteria of more than twenty percent of pipe circumference of fifty percent of through-wall depth. In the future, we intend to perform flow sizing with automated UT techniques in addition to manual sizing to provide more information for comparison with future examinations.

  2. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    Science.gov (United States)

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  3. Water treatment process in the JEN-1 Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Urgel, M.; Perez-Bustamante, J. A.; Batuecas, T.

    1965-07-01

    The main characteristics and requirements which must be met with by waters to be used for nuclear reactors were studied paying attention separately both to those used in primary and secondary circuits as well as to the purification systems to be employed in each case. The experiments carried out for the initial pretreatment of water and the ion-exchange de ionization processes including a number of systems consisting of separated and mixed beds loaded with a variety of different commercially available resins are described. (Author) 24 refs.

  4. Water to atmosphere fluxes of {sup 131}I in relation with alkyl-iodide compounds from the Seine Estuary (France)

    Energy Technology Data Exchange (ETDEWEB)

    Connan, Olivier [Laboratoire de Radioecologie de Cherbourg-Octeville, Institut de Radioprotection et de Surete nucleaire (IRSN), Service d' Etudes et du Comportement des Radionucleides dans l' Environnement (SECRE), rue Max Pol Fouchet, 50130 Cherbourg-Octeville (France)], E-mail: olivier.connan@irsn.fr; Tessier, Emmanuel [Laboratoire de Chimie Analytique et Bio-Inorganique et Environnement, UMR CNRS universite de Pau et des Pays de l' Adour, Helioparc Pau Pyrenees, 2 Avenue Pierre Angot, 64053 Pau Cedex 9 (France); Maro, Denis [Laboratoire de Radioecologie de Cherbourg-Octeville, Institut de Radioprotection et de Surete nucleaire (IRSN), Service d' Etudes et du Comportement des Radionucleides dans l' Environnement (SECRE), rue Max Pol Fouchet, 50130 Cherbourg-Octeville (France); Amouroux, David [Laboratoire de Chimie Analytique et Bio-Inorganique et Environnement, UMR CNRS universite de Pau et des Pays de l' Adour, Helioparc Pau Pyrenees, 2 Avenue Pierre Angot, 64053 Pau Cedex 9 (France); Hebert, Didier; Rozet, Marianne; Voiseux, Claire; Solier, Luc [Laboratoire de Radioecologie de Cherbourg-Octeville, Institut de Radioprotection et de Surete nucleaire (IRSN), Service d' Etudes et du Comportement des Radionucleides dans l' Environnement (SECRE), rue Max Pol Fouchet, 50130 Cherbourg-Octeville (France)

    2008-07-15

    This study presents an original work on measurements of stable and radioactive iodinated species in the Seine estuary (France), with estimates fluxes of volatile gaseous species from water to the atmosphere. Various iodinated compounds were identified in water and air in particular {sup 131}I in water, what is unusual. Concentrations and behaviour of iodinated elements in the Seine estuary seem similar to what has been observed in other European estuaries. MeI (Methyl Iodide) and Total Volatile Iodine (TVI) fluxes from water to air vary between 392 and 13949 pmol m{sup -2} d{sup -1} and between 1279 and 16484 pmol m{sup -2} d{sup -1}, respectively. Water to air flux of TVI for the Seine river was estimated in the range 4-46 kg y{sup -1}. Measurements of {sup 131}I in water varying between 0.4 and 11.9 Bq m{sup -3}. Fluxes of {sup 131}I from water to atmosphere are in the range 2.4 x 10{sup 5}-1.3 x 10{sup 7} Bq y{sup -1}, close to an annual discharge of {sup 131}I by a nuclear reactor.

  5. Simulation of Water Gas Shift Zeolite Membrane Reactor

    Science.gov (United States)

    Makertiharta, I. G. B. N.; Rizki, Z.; Zunita, Megawati; Dharmawijaya, P. T.

    2017-07-01

    The search of alternative energy sources keeps growing from time to time. Various alternatives have been introduced to reduce the use of fossil fuel, including hydrogen. Many pathways can be used to produce hydrogen. Among all of those, the Water Gas Shift (WGS) reaction is the most common pathway to produce high purity hydrogen. The WGS technique faces a downstream processing challenge due to the removal hydrogen from the product stream itself since it contains a mixture of hydrogen, carbon dioxide and also the excess reactants. An integrated process using zeolite membrane reactor has been introduced to improve the performance of the process by selectively separate the hydrogen whilst boosting the conversion. Furthermore, the zeolite membrane reactor can be further improved via optimizing the process condition. This paper discusses the simulation of Zeolite Membrane Water Gas Shift Reactor (ZMWGSR) with variation of process condition to achieve an optimum performance. The simulation can be simulated into two consecutive mechanisms, the reaction prior to the permeation of gases through the zeolite membrane. This paper is focused on the optimization of the process parameters (e.g. temperature, initial concentration) and also membrane properties (e.g. pore size) to achieve an optimum product specification (concentration, purity).

  6. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  7. Transactions of the nineteenth water reactor safety information meeting

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. (comp.)

    1991-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.

  8. Non-linear analysis in Light Water Reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  9. Dominant accident sequences in Oconee-1 pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dearing, J F; Henninger, R J; Nassersharif, B

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling.

  10. Materials Inventory Database for the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  11. The Water Suitcase of Migrants: Assessing Virtual Water Fluxes Associated to Human Migration.

    Directory of Open Access Journals (Sweden)

    Rodolfo Metulini

    Full Text Available Disentangling the relations between human migrations and water resources is relevant for food security and trade policy in water-scarce countries. It is commonly believed that human migrations are beneficial to the water endowments of origin countries for reducing the pressure on local resources. We show here that such belief is over-simplistic. We reframe the problem by considering the international food trade and the corresponding virtual water fluxes, which quantify the water used for the production of traded agricultural commodities. By means of robust analytical tools, we show that migrants strengthen the commercial links between countries, triggering trade fluxes caused by food consumption habits persisting after migration. Thus migrants significantly increase the virtual water fluxes and the use of water in the countries of origin. The flux ascribable to each migrant, i.e. the "water suitcase", is found to have increased from 321 m3/y in 1990 to 1367 m3/y in 2010. A comparison with the water footprint of individuals shows that where the water suitcase exceeds the water footprint of inhabitants, migrations turn out to be detrimental to the water endowments of origin countries, challenging the common perception that migrations tend to relieve the pressure on the local (water resources of origin countries.

  12. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    Directory of Open Access Journals (Sweden)

    Weimin Ma

    2016-03-01

    Full Text Available A historical review of in-vessel melt retention (IVR is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs. The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

  13. Candidate materials performance under Supercritical Water Reactor (SCWR) conditions

    Energy Technology Data Exchange (ETDEWEB)

    Toivonen, A.; Penttilae, S.; Rissanen, L. (VTT Technical Research Centre of Finland, Espoo (Finland))

    2010-05-15

    The High Performance Light Water Reactor (HPLWR) is working at super-critical pressure (25 MPa) and a core coolant temperature up to 500 deg C. As an evolutionary step this reactor type follows the development path of modern supercritical coal-fired plants. This paper reviews the results on performance of commercial candidate materials for in-core applications focusing on corrosion, stress corrosion cracking (SCC) and creep issues. General corrosion (oxidation) tests with an inlet oxygen concentration of 125-150 ppb have been performed on several iron and nickel alloys at 300 to 650 deg C and 25 MPa in supercritical water. Stress corrosion cracking (SCC) susceptibility of selected austenitic stainless steels and a high chromium ODS (Oxide Dispersion Strengthened) alloy were also studied in slow strain rate tests (SSRT) in supercritical water at 500 deg C and 650 deg C. Furthermore, constant load creep tests have been performed on selected austenitic steels at 500 deg C and 650 deg C in supercritical water (25 MPa, 1 ppm O{sub 2}) and in an inert atmosphere (He, pressure 1 atm). Based on the materials studies, the current candidate materials for the core internals are austenitic steels with sufficient oxidation and creep resistance up to 500-550 deg C. High chromium austenitic steels and ODS alloys steels are considered for the fuel rod cladding due to their oxidation resistance up to 650 deg C. However, problems with manufacturing and joining of ODS alloys need to be solved. Alloys with high nickel content were not considered for the SCC or creep studies because of the strong effect of Ni on neutronics of the reactor core (orig.)

  14. Residual fluxes of water, salt and suspended sediment in the Beypore Estuary

    Digital Repository Service at National Institute of Oceanography (India)

    AnilKumar, N.; Revichandran, C.; Sankaranarayanan, V.N.; Josanto, V.

    The monthly trends of the residual fluxes of salt and water and the transportation of suspended sediments in the Beypore estuarine system, Kerala, India were examined. At the river mouth the water flux was directed seaward during the postmonsoon...

  15. Development of neutron measurement techniques in reactor diagnostics and determination of water content and water flow

    Energy Technology Data Exchange (ETDEWEB)

    Avdic, Senada

    2000-09-01

    The present thesis deals with three comparatively different topics in neutron physics research. These topics are as follows: construction and experimental investigation of a new detector, capable of measuring the neutron current, and investigation of the possibility to use it for the localisation of a neutron source in a simple experimental arrangement; execution of neutron transmission measurements based on a stationary neutron generator, and the study of their suitability for determining the volume porosity of geological samples; study of the possibility for improving the accuracy of water flow measurements based on the pulsed neutron activation technique. The first subject of this thesis concerns the measurement of the neutron current by a newly constructed detector. The motivation for this work stems from a recent suggestion that the performance of core monitoring methods could be enhanced if, in addition to the scalar neutron flux, also the neutron current was measured. To this end, a current detector was based on a scintillator mounted on a fibre and a Cd layer on one side of the detector. The measurements of the 2-D neutron current were performed in an experimental system by using this detector. The efficiency of the detector in reactor diagnostics was illustrated by demonstrating that the position of a neutron source can be determined by measuring the scalar neutron flux and the neutron current in one spatial point. The results of measurement and calculation show both the suitability of the detector construction for the measurement of the neutron current vector and the use of the current in diagnostics and monitoring. The second subject of this thesis concerns fast neutron transmission measurements, based on a stationary neutron generator, for determining the volume porosity of a sample in a model experiment. Such a technique could be used in field measurements with obvious advantages in comparison with thermal neutron transmission techniques, which can

  16. Drinking water treatment using a submerged internal-circulation membrane coagulation reactor coupled with permanganate oxidation.

    Science.gov (United States)

    Zhang, Zhongguo; Liu, Dan; Qian, Yu; Wu, Yue; He, Peiran; Liang, Shuang; Fu, Xiaozheng; Li, Jiding; Ye, Changqing

    2017-06-01

    A submerged internal circulating membrane coagulation reactor (MCR) was used to treat surface water to produce drinking water. Polyaluminum chloride (PACl) was used as coagulant, and a hydrophilic polyvinylidene fluoride (PVDF) submerged hollow fiber microfiltration membrane was employed. The influences of trans-membrane pressure (TMP), zeta potential (ZP) of the suspended particles in raw water, and KMnO 4 dosing on water flux and the removal of turbidity and organic matter were systematically investigated. Continuous bench-scale experiments showed that the permeate quality of the MCR satisfied the requirement for a centralized water supply, according to the Standards for Drinking Water Quality of China (GB 5749-2006), as evaluated by turbidity (water flux, the removal of turbidity, TOC and dissolved organic carbon (DOC) in the raw water also increased with increasing TMP in the range of 0.01-0.05MPa. High ZP induced by PACl, such as 5-9mV, led to an increase in the number of fine and total particles in the MCR, and consequently caused serious membrane fouling and high permeate turbidity. However, the removal of TOC and DOC increased with increasing ZP. A slightly positive ZP, such as 1-2mV, corresponding to charge neutralization coagulation, was favorable for membrane fouling control. Moreover, dosing with KMnO 4 could further improve the removal of turbidity and DOC, thereby mitigating membrane fouling. The results are helpful for the application of the MCR in producing drinking water and also beneficial to the research and application of other coagulation and membrane separation hybrid processes. Copyright © 2016. Published by Elsevier B.V.

  17. Water flux in animals: analysis of potential errors in the tritiated water method

    Energy Technology Data Exchange (ETDEWEB)

    Nagy, K.A.; Costa, D.

    1979-03-01

    Laboratory studies indicate that tritiated water measurements of water flux are accurate to within -7 to +4% in mammals, but errors are larger in some reptiles. However, under conditions that can occur in field studies, errors may be much greater. Influx of environmental water vapor via lungs and skin can cause errors exceeding +-50% in some circumstances. If water flux rates in an animal vary through time, errors approach +-15% in extreme situations, but are near +-3% in more typical circumstances. Errors due to fractional evaporation of tritiated water may approach -9%. This error probably varies between species. Use of an inappropriate equation for calculating water flux from isotope data can cause errors exceeding +-100%. The following sources of error are either negligible or avoidable: use of isotope dilution space as a measure of body water volume, loss of nonaqueous tritium bound to excreta, binding of tritium with nonaqueous substances in the body, radiation toxicity effects, and small analytical errors in isotope measurements. Water flux rates measured with tritiated water should be within +-10% of actual flux rates in most situations.

  18. The Development of the Advanced Light Water Reactor in Korea - The Korean Next Generation Reactor -

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H.G.; Park, Y.S.; Kim, B.S.; Cho, S.J. [Korea Electric Power Research Institute, Taejon (Korea)

    1999-07-01

    Korean next generation reactor (KNGR), which is to be designed as a standard evolutionary advanced light water reactor (ALWR) in Korea, has been developed from 1992 as one of long-term government projects. The major characteristics of the KNGR are as follows; KNGR is 2-loop PWR and its design lift time is 60 years. The CDF and the CFF will be much lower than 10{sup -5}/RY and 10{sup -6}/RY, respectively. For the design improvement, KNGR adopted inconel-690 as a steam generator tube material, four train ECCS, refueling water storage tank inside containment, and double cylindrical concrete containment. For more reliable and easier control, compact workstations have been adopted in the design of main control complex and digital I and C technology is used for protection, control, and monitoring. In addition, KNGR has some passive design features such as fluidic device in safety injection tank, passive secondary condensing system, and passive auto-catalytic hydrogen recombiner to enhance safety. (author). 4 refs., 4 figs.

  19. Aspects of the physics and chemistry of water radiolysis by fast neutrons and fast electrons in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCracken, D.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Tsang, K.T. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Laughton, P.J

    1998-09-01

    Detailed radiation physics calculations of energy deposition have been done for the coolant of CANDU reactors and Pressurized Water Reactors (PWRs). The geometry of the CANDU fuel channel was modelled in detail. Fluxes and energy-deposition rates for neutrons, recoil ions, photons, and fast electrons have been calculated using MCNP4B, WIMS-AECL, and specifically derived energy-transfer factors. These factors generate the energy/flux spectra of recoil ions from fast-neutron energy/flux spectra. The energy spectrum was divided into 89 discrete ranges (energy bins).The production of oxidizing species and net coolant radiolysis can be suppressed by the addition of hydrogen to the coolant of nuclear reactors. It is argued that the net dissociation of coolant by gamma rays is suppressed by lower levels of excess hydrogen than when dissociation is by ion recoils. This has consequences for the modelling of coolant radiolysis by homogeneous kinetics. More added hydrogen is required to stop water radiolysis by recoil ions acting alone than if recoil ions and gamma rays acted concurrently in space and time. Homogeneous kinetic models and experimental data suggest that track overlap is very inefficient in providing radicals from gamma-ray tracks to recombine molecular products in ion-recoil tracks. An inhomogeneous chemical model is needed that incorporates ionizing-particle track structure and track overlap. Such a model does not yet exist, but a number of limiting cases using homogeneous kinetics are discussed. There are sufficient uncertainties and contradictions in the data relevant to the radiolysis of reactor coolant that the relatively high CHC's (critical hydrogen concentration) observed in NRU reactor experiments (compared to model predictions) may be explainable by errors in fundamental data and understanding of water radiolysis under reactor conditions. The radiation chemistry program at CRL has been focused to generate quantitative water-radiolysis data in a

  20. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  1. Experiences with austenitic steels in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wachter, O. [Preussische Elektrizitaets-AG (Preussenelektra), Hannover (Germany); Bruemmer, G. [Hamburgische Electricitaets-Werke AG., Hamburg (Germany)

    1997-05-01

    Stabilized austenitic steels are susceptible to intergranular stress corrosion cracking (IGSCC) under boiling water reactor (BWR) conditions. This important finding for the German nuclear power station industry arises from the detection of cracks during the last 3 years in reactor hot water pipes made from titanium-stabilized steel AISI 321 in six BWRs and in reactor core components made from the niobium-stabilized steel AISI 347 in one BWR. All the observed cracks had a common feature: they had their origin in the chromium carbide precipitates at the grain boundaries and in the associated chromium-depleted region near the grain boundary. These microstructural features in the heat-affected zones of the hot water pipe weldments were caused by the heat input during deposition of the root bead. The TiC partially dissolved in the region near the fusion line and the released carbon reacted to form chromium-rich M{sub 23}C{sub 6}. Regarding the cracks found in the core shroud and the core grid plates, it was shown that a sensitizing heat treatment of rings taken from the same heat of steel could give rise to a microstructure susceptible to IGSCC in the region of a weldment. High carbon contents coupled with low stabilization ratios led to sensitization. Residual stresses developed during welding provided the significant contributions to the tensile stress necessary for IGSCC. With regard to the service medium, the influence of the electrochemical corrosion potential (ECP) was recognized as a dominant factor, together with the conductivity. The corrosion potential was mainly determined by the radiolytic formation of H{sub 2}O{sub 2}; with increasing distance from the core, the H{sub 2}O{sub 2} content decreased owing to catalytic decomposition. For the pipes the problem of IGSCC could be resolved by the use of optimized steel (lower carbon content with maximum allowable stabilization ratio).

  2. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  3. The radiochemistry of nuclear power plants with light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neeb, K.H.

    1997-12-31

    In this book, radioactivity and the chemical reactions of radionuclides within the different areas of a nucler power plant are discussed. The text concentrates on commercially operated light water reactors which currently represent the greatest fraction by far of the world`s nuclear power capacity. This book is not only intended for experts working in the various fields of radiochemistry in nuclear power plants. It also provides an overview of the topics dealt with for the operators of nuclear power plants, for people working in design and development and safety-related areas, as well as for those working in licensing and supervision. (orig.)

  4. Advanced fuels for plutonium management in pressurized water reactors

    Science.gov (United States)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

  5. Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hurt, Christopher J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hobbs, Randy W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (238Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (237Np) dioxide (NpO2)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs

  6. Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Manglik, R M; Athavale, A; Kalaikadal, D S; Deodhar, A; Verma, U

    2011-09-02

    The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

  7. Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

    Science.gov (United States)

    Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

    2017-02-01

    Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

  8. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  9. Platform for monitoring water and solid fluxes in mountainous rivers

    Science.gov (United States)

    Nord, Guillaume; Esteves, Michel; Aubert, Coralie; Belleudy, Philippe; Coulaud, Catherine; Bois, Jérôme; Geay, Thomas; Gratiot, Nicolas; Legout, Cédric; Mercier, Bernard; Némery, Julien; Michielin, Yoann

    2016-04-01

    The project aims to develop a platform that electronically integrates a set of existing sensors for the continuous measurement at high temporal frequency of water and solid fluxes (bed load and suspension), characteristics of suspended solids (distribution in particle size, settling velocity of the particles) and other variables on water quality (color, nutrient concentration). The project is preferentially intended for rivers in mountainous catchments draining areas from 10 to 1000 km², with high suspended sediment concentrations (maxima between 10 and 300 g/l) and highly dynamic behavior, water discharge varying of several orders of magnitude in a short period of time (a few hours). The measurement of water and solid fluxes in this type of river remains a challenge and, to date, there is no built-in device on the market to continuously monitor all these variables. The development of this platform is based on a long experience of measurement of sediment fluxes in rivers within the French Critical Zone Observatories (http://portailrbv.sedoo.fr/), especially in the Draix-Bléone (http://oredraixbleone.irstea.fr/) and OHMCV (http://www.ohmcv.fr/) observatories. The choice was made to integrate in the platform instruments already available on the market and currently used by the scientific community (water level radar, surface velocity radar, turbidity sensor, automatic water sampler, video camera) and to include also newly developed instruments (System for the Characterization of Aggregates and Flocs - see EGU2016-8542 - and hydrophone) or commercial instruments (spectrophotometer and radiometer) to be tested in surface water with high suspended sediment concentration. Priority is given to non-intrusive instruments due to their robustness in this type of environment with high destructive potential. Development work includes the construction of a platform prototype "smart" and remotely configurable for implantation in an isolated environment (absence of electric

  10. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  11. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    Energy Technology Data Exchange (ETDEWEB)

    Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

    2014-07-01

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. The paper describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motorgenerator set.

  12. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Paul Y [Los Alamos National Laboratory

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-08-10

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

  14. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  15. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  16. Measurement of in-phantom neutron flux and gamma dose in Tehran research reactor boron neutron capture therapy beam line.

    Science.gov (United States)

    Bavarnegin, Elham; Sadremomtaz, Alireza; Khalafi, Hossein; Kasesaz, Yaser

    2016-01-01

    Determination of in-phantom quality factors of Tehran research reactor (TRR) boron neutron capture therapy (BNCT) beam. The doses from thermal neutron reactions with 14N and 10B are calculated by kinetic energy released per unit mass approach, after measuring thermal neutron flux using neutron activation technique. Gamma dose is measured using TLD-700 dosimeter. Different dose components have been measured in a head phantom which has been designed and constructed for BNCT purpose in TRR. Different in-phantom beam quality factors have also been determined. This study demonstrates that the TRR BNCT beam line has potential for treatment of superficial tumors.

  17. Reactor vessel water level estimation during severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Yoo, Kwae Hwan; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2016-06-15

    Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured. The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks. The developed CFNN model was found to be sufficiently accurate for estimating the reactor vessel water level when the sensor performance had deteriorated. Therefore, the developed CFNN model can help provide effective information to operators in the event of a severe accident.

  18. Improvement of Critical Heat Flux Correlation for Research Reactors using Plate-Type Fuel

    National Research Council Canada - National Science Library

    KAMINAGA, Masanori; YAMAMOTO, Kazuyoshi; SUDO, Yukio

    1998-01-01

    ... reversal.The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition...

  19. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of fuel and claddings during accident are still below limitations which are in secure condition.

  20. The application of noble metals in light-water reactors

    Science.gov (United States)

    Kim, Young-Jin; Niedrach, Leonard W.; Indig, Maurice E.; Andresen, Peter L.

    1992-04-01

    Corrosion potential is a primary determinant of the stress-corrosion cracking susceptibility of structural materials in high-temperature water. Efforts to minimize stress-corrosion cracking in light-water reactors include adding hydrogen. In someplants' out-of-core regions, the hydrogen required to achieve the desired corrosion potential is relatively high. In-core, more hydrogen is needed for an equivalent reduction in corrosion potential. Additionally, sIDe effects of high hydrogen-addition rates, including increased 16N turbine shine and 60CO deposition, have also been observed in some cases. An approach involving noble-metal coatings on and alloying additions to engineering materials dramatically improves the efficiency with which the corrosion potential is decreased as a function of hydrogen addition, such that very low potentials are obtained once a stoichiometric concentration of hydrogen (versus oxygen) is achieved.

  1. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  2. Fluxes of water and energy in physically heterogeneous environments

    Energy Technology Data Exchange (ETDEWEB)

    Breshears, D.D.; Barnes, F.J.; Davenport, D.W. [and others

    1997-11-01

    This is the final report of a one-year, Laboratory Directed Research and Development (LDRD) project at Los Alamos National Laboratory (LANL). Fluxes of water and energy at the near-surface environment are tightly interrelated with a heterogeneous vegetation pattern that is a mosaic of tree canopies and intercanopy area. The objective was to improve the ability to predict these interrelationships, which are not well quantified. The authors (1) quantified how vegetation overstory determines the patterns of soil moisture and near-ground solar radiation, (2) developed spatial neighborhood analyses that demonstrated how woody plants exploit canopy/intercanopy heterogeneity, (3) developed a spatially explicit model for predicting near-ground solar radiation for sites along a grassland-forest continuum, (4) developed a water balance model that predicted temporal shifts in soil moisture between canopy and intercanopy patches, and (5) used the collective results to assess large-scale ecosystem responses to climate variations that lead to accelerated soil erosion.

  3. Surface fluxes and water balance of spatially varying vegetation within a small mountainous headwater catchment

    Science.gov (United States)

    Understanding the role of ecosystems in modulating energy, water and carbon fluxes is critical to quantifying the variability in energy, carbon, and water balances across landscapes. This study compares and contrasts the seasonal surface fluxes of sensible heat, latent heat and carbon fluxes measur...

  4. Design and Thermal Analysis for Irradiation of Pyrolytic Carbon/Silicon Carbide Diffusion Couples in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept consisting of a spherical fuel kernel made of uranium oxide and uranium carbide, surrounded by a porous carbonaceous buffer layer and successive layers of dense inner pyrolytic carbon (IPyC), silicon carbide (SiC) deposited by chemical vapor , and dense outer pyrolytic carbon (OPyC). This fuel concept is being considered for advanced reactor applications such as high temperature gas-cooled reactors (HTGRs) and molten salt reactors (MSRs), as well as for accident-tolerant fuel for light water reactors (LWRs). Development and implementation of TRISO fuel for these reactor concepts support the US Department of Energy (DOE) Office of Nuclear Energy mission to promote safe, reliable nuclear energy that is sustainable and environmentally friendly. During operation, the SiC layer serves as the primary barrier to metallic fission products and actinides not retained in the kernel. It has been observed that certain fission products are released from TRISO fuel during operation, notably, Ag, Eu, and Sr [1]. Release of these radioisotopes causes safety and maintenance concerns.

  5. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  6. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Busby, Jeremy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barnard, Cathy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  7. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  8. Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States); Peko, D. [US Dept. of Energy, Washington, DC (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Humrickhouse, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  9. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

    2015-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  10. Stress analysis of the reactor pressure vessel of the high performance light water reactors (HPLWR); Festigkeitsanalyse fuer den Reaktordruckbehaelter des High Performance Light Water Reactor (HPLWR)

    Energy Technology Data Exchange (ETDEWEB)

    Guelton, E.; Fischer, K.

    2006-12-15

    The High Performance Light Water Reactor (HPLWR) is one of the concepts of the Generation IV program. The main difference compared to current Light Water Reactors (LWR) results from the supercritical steam condition of the coolant. Due to the supercritical pressure of 25 MPa, water, used as moderator and coolant, flows as a single phase through the core. The temperatures at the outlet are above 500 C. These conditions have a major impact on the design of the Reactor Pressure Vessel (RPV). For the modelling a RPV concept is proposed, which resembles the design of current LWR and allows the use of approved materials on one side and also meets the additional demands on the other side. A first dimensioning of the RPV wall thicknesses and the geometrical proportions has been performed using the german KTA-guidelines. To verify these results, a stress analysis using the finite element method has been performed with the program ANSYS. The combined mechanical and thermal calculations provide the primary, secondary and peak stresses which are evaluated using the KTA-guidelines design loading (Level 0) and service loading level A for the different components. The results confirm the wall thicknesses estimated by Fischer et al. (2006), but there are peak stresses in the vicinity of the inlet and outlet flanges, which are very close to the allowed design limit. For larger diameters of the RPV those regions will become critical and the stresses might exceed the design limits. Design optimizations for those regions are proposed and evaluated. A readjusted geometry of the inlet flange reduces those stresses by 65%. (orig.)

  11. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used.

  12. Benchmarking severe accident computer codes for heavy water reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J.H. [International Atomic Energy Agency, Vienna (Austria)

    2010-07-01

    Consideration of severe accidents at a nuclear power plant (NPP) is an essential component of the defence in depth approach used in nuclear safety. Severe accident analysis involves very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. International cooperative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through coordinated research projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes. The IAEA is organizing a CRP on benchmarking severe accident computer codes for heavy water reactor (HWR) applications. The CRP scope includes defining the severe accident sequence and conducting benchmark analyses for HWRs, evaluating the capabilities of existing computer codes to predict important severe accident phenomena, and suggesting necessary code improvements and/or new experiments to reduce uncertainties. The CRP has been planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for HWRs. (author)

  13. Controlling radiation fields in siemans designed light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riess, R.; Marchl, T. [Siemens Power Generation Group, Erlangen (Germany)

    1995-03-01

    An essential item for the control of radiation fields is the minimization of the use of satellites in the reactor systems of Light Water Reactors (LWRs). A short description of the qualification of Co-replacement materials will be followed by an illustration of the locations where these materials were implemented in Siemens designed LWRs. Especially experiences in PWRs show the immense influence of reduction of cobalt sources on dose rate buildup. The corrosion and the fatique and wear behavior of the replacement materials has not created concern up to now. A second tool to keep occupational radiation doses at a low level in PWRs is the use of the modified B/Li-chemistry. This is practized in Siemens designed plants by keeping the Li level at a max. value of 2 ppm until it reaches a pH (at 300{degrees}C) of {approximately}7.4. This pH is kept constant until the end of the cycle. The substitution of cobalt base alloys and thus the removal of the Co-59 sources from the system had the largest impact on the radiation levels. Nonetheless, the effectiveness of the coolant chemistry should not be neglected either. Several years of successful operation of PWRs with the replacement materials resulted in an occupational radiation exposure which is below 0.5 man-Sievert/plant and year.

  14. Technologies for Upgrading Light Water Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  15. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  16. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  17. A study on the establishment of component/equipment performance criteria considering Heavy Water Reactor characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Kwon, Young Chul; Lee, Min Kyu; Lee, Yun Soo [Sunmoon Univ., Asan (Korea, Republic of); Chang, Seong Hoong; Ryo, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Soong Pyung; Hwnag, Jung Rye; Chung, Chul Kee [Chosun Univ., Gwangju (Korea, Republic of)

    2002-03-15

    Foreign and domestic technology trends, regulatory requirements, design and researches for heavy water reactors are analyzed. Safety design guides of Canada industry and regulatory documents and consultative documents of Canada regulatory agency are reviewed. Applicability of MOST guidance 16 Revision 'guidance for technical criteria of nuclear reactor facility' is reviewed. Specific performance criteria are established for components and facilities for heavy water reactor.

  18. Application of Forward Osmosis Membrane in a Sequential Batch Reactor for Water Reuse

    KAUST Repository

    Li, Qingyu

    2011-07-01

    Forward osmosis (FO) is a novel membrane process that potentially can be used as an energy-saving alternative to conventional membrane processes. The objective of this study is to investigate the performance of a FO membrane to draw water from wastewater using seawater as draw solution. A study on a novel osmotic sequential batch reactor (OsSBR) was explored. In this system, a plate and frame FO cell including two flat-sheet FO membranes was submerged in a bioreactor treating the wastewater. We found it feasible to treat the wastewater by the OsSBR process. The DOC removal rate was 98.55%. Total nitrogen removal was 62.4% with nitrate, nitrite and ammonium removals of 58.4%, 96.2% and 88.4% respectively. Phosphate removal was almost 100%. In this OsSBR system, the 15-hour average flux for a virgin membrane with air scouring is 3.103 LMH. After operation of 3 months, the average flux of a fouled membrane is 2.390 LMH with air scouring (23% flux decline). Air scouring can help to remove the loose foulants on the active layer, thus helping to maintain the flux. Cleaning of the FO membrane fouled in the active layer was probably not effective under the conditions of immersing the membrane in the bioreactor. LC-OCD results show that the FO membrane has a very good performance in rejecting biopolymers, humics and building blocks, but a limited ability in rejecting low molecular weight neutrals.

  19. Construction management of Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bohra, S.A. [Nuclear Power Corporation of India Limited, Vikram Sarabhai Bhavan, Anushaktinagar, Mumbai 400094 (India)]. E-mail: sabohra@npcil.co.in; Sharma, P.D. [Nuclear Power Corporation of India Limited, Vikram Sarabhai Bhavan, Anushaktinagar, Mumbai 400094 (India)

    2006-04-15

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  20. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  1. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  2. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact... of Title 10 of the Code of Federal Regulations (10 CFR) for the La Crosse Boiling Water Reactor...

  3. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI AGENCY...) 73.55, for the LaCrosse Boiling Water Reactor (LACBWR). This Environmental Assessment (EA) has been...

  4. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  5. Final Report on Isotope Ratio Techniques for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

    2009-07-01

    The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

  6. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  7. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  8. Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

    Science.gov (United States)

    Carcreff, H.; Salmon, L.; Lepeltier, V.; Guyot, J. M.; Bouard, E.

    2018-01-01

    Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the "zero method". Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the "zero method" measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between "zero" and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions, obtained with the Self-Power Neutron

  9. Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

    Directory of Open Access Journals (Sweden)

    Carcreff H.

    2018-01-01

    Full Text Available Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the “zero method”. Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the “zero method” measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between “zero” and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions

  10. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  11. Light Water Reactor Sustainability Program: Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-05-01

    proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

  12. Procedure of calculation of the spatial distribution of temperatures and heat fluxes in the steam generator of a nuclear power installation with an RBEC fast-neutron reactor

    Science.gov (United States)

    Frolov, A. A.; Sedov, A. A.

    2016-08-01

    A method for combined 3D/1D-modeling of thermohydraulics of a once-through steam generator (SG) based on the joint analysis of three-dimensional thermo- and hydrodynamics of a single-phase heating coolant in the intertube space and one-dimensional thermohydraulics of steam-generating channels (tubes) with the use of well-known friction and heat-transfer correlations under various boiling conditions is discussed. This method allows one to determine the spatial distribution of temperatures and heat fluxes of heat-exchange surfaces of SGs with a single-phase heating coolant in the intertube space and with steam generation within tubes. The method was applied in the analytical investigation of typical operation of a once-through SG of a nuclear power installation with an RBEC fast-neutron heavy-metal reactor that is being designed by Kurchatov Institute in collaboration with OKB GIDROPRESS and Leipunsky Institute of Physics and Power Engineering. Flow pattern and temperature fields were obtained for the heavy-metal heating coolant in the intertube space. Nonuniformities of heating of the steam-water coolant in different heat-exchange tubes and nonuniformities in the distribution of heat fluxes at SG heat-exchange surfaces were revealed.

  13. Full-Scale Modeling Explaining Large Spatial Variations of Nitrous Oxide Fluxes in a Step-Feed Plug-Flow Wastewater Treatment Reactor.

    Science.gov (United States)

    Ni, Bing-Jie; Pan, Yuting; van den Akker, Ben; Ye, Liu; Yuan, Zhiguo

    2015-08-04

    Nitrous oxide (N2O) emission data collected from wastewater treatment plants (WWTPs) show huge variations between plants and within one plant (both spatially and temporarily). Such variations and the relative contributions of various N2O production pathways are not fully understood. This study applied a previously established N2O model incorporating two currently known N2O production pathways by ammonia-oxidizing bacteria (AOB) (namely the AOB denitrification and the hydroxylamine pathways) and the N2O production pathway by heterotrophic denitrifiers to describe and provide insights into the large spatial variations of N2O fluxes in a step-feed full-scale activated sludge plant. The model was calibrated and validated by comparing simulation results with 40 days of N2O emission monitoring data as well as other water quality parameters from the plant. The model demonstrated that the relatively high biomass specific nitrogen loading rate in the Second Step of the reactor was responsible for the much higher N2O fluxes from this section. The results further revealed the AOB denitrification pathway decreased and the NH2OH oxidation pathway increased along the path of both Steps due to the increasing dissolved oxygen concentration. The overall N2O emission from this step-feed WWTP would be largely mitigated if 30% of the returned sludge were returned to the Second Step to reduce its biomass nitrogen loading rate.

  14. Studies of low temperature, low flux radiation embrittlement of nuclear reactor structural materials. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Lucas, G.E.

    1993-06-01

    There are several existing research programs which have components pertinent to the issue of low flux/low temperature embrittlement; in particular, examination of the Shippingport shield tank which has been exposed to low flux and relatively low temperature is being performed by ANL, and evaluation of low temperature embrittlement in A508 and A533B steels in support of the HTGR is currently being performed by ORNL. However, these programs are not specifically directed at the broader issue of low flux/low temperature embrittlement in a range of structural steels. Hence, the authors coordinated their effort with these programs so that their investigations were complementary to existing programs, and they focused on a set of materials which expand the data base developed in these programs. In particular, the authors have investigated embrittlement phenomena in steels that are similar to those used in support structure.

  15. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    Science.gov (United States)

    Lasche, George P.

    1988-01-01

    A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

  16. Stability analysis of the high performance light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ortega Gomez, Tino

    2009-03-15

    In the Generation IV international advanced nuclear reactor development program, the High Performance Light Water Reactor (HPLWR) is one of the most promising candidates. Important features are its inherently high thermodynamic efficiency (of approximately 45 %) and the ability to use existing supercritical water technology which previously has been developed and deployed for fossil fired power plants. Within a HPLWR core, the fluid experiences a drastic change in thermal and transport properties such as density, dynamic viscosity, specific heat and thermal conductivity, as the supercritical water is heated from 280 C to 500 C. The density change substantially exceeds that in a Boiling Water Reactor (i.e., HPLWR: density changes from 780 kg/m{sup 3} to 90 kg/m{sup 3}; BWR: density changes from 750 kg/m{sup 3} to 198 kg/m3). Due to this density change, the HPLWR can be - under certain operation parameters - susceptible to various thermal-hydraulic flow instabilities, which have to be avoided. In this thesis a stability analysis for the HPLWR is presented. This analysis is based on analytical considerations and numerical results, which were obtained by a computer code developed by the author. The heat-up stages of the HPLWR three-pass core are identified in respect to the relevant flow instability phenomena. The modeling approach successfully used for BWR stability analysis is extended to supercritical pressure operation conditions. In particular, a one-dimensional equation set representing the coolant flow of HPLWR fuel assemblies has been implemented in a commercial software named COMSOL to perform steady-state, time-dependent, and modal analyses. An investigation of important static instabilities (i.e., Ledinegg instabilities, flow maldistribution) and Pressure Drop Oscillations (PDO) have been carried out and none were found under operation conditions of the HPLWR. Three types of Density Wave Oscillation (DWO) modes have been studied: the single channel DWO, the

  17. Design and Fluid Dynamic Investigations for a High Performance Light Water Reactor Fuel Assembly

    Science.gov (United States)

    Hofmeister, Jan; Laurin, Eckart; Class, Andreas G.

    2005-11-01

    Within the 5th Framework Program of the European Commission a nuclear light water reactor with supercritical steam conditions has been investigated called High Performance Light Water Reactor (HPLWR). This reactor concept is distinct from conventional light water reactor concepts by the fact, that supercritical water is used to achieve higher core outlet temperatures. The reactor operates with a high system pressure, high heat-up of the coolant within the core, and high outlet temperatures of the coolant resulting in a thermal efficiency of up to 44%. We present the design concept proposed by IKET, and a fluid dynamic problem in the foot piece of the fuel assembly, where unacceptable temperature variations must be omitted.

  18. Theoretical studies on the new system of neutron flux measurement in the nuclear reactor

    CERN Document Server

    Wan Jun Sheng; Zhao Zhu Min; Chen Li Xin; Jing Chun Yuan; Wang Dao Hua

    2002-01-01

    The neutron measurement system based on the Nuclear Pumped Laser (NPL) techniques was explored in this work utilizing the relationship between the neutron flux and the NPL laser power. Following the studies on the mechanism of the NPL, the laser intrinsic efficiency and the energy deposition efficiency in the laser cell of this system were theoretically studied in detail. A laser pumping mechanism was proposed. The variation of sensitivity with working time and the response function on neutron flux was discussed. The feasibility of this neutron measurement system was demonstrated theoretically

  19. Absolute measurement of D-T neutron flux with a monitor using activation of flowing water

    Energy Technology Data Exchange (ETDEWEB)

    Uno, Y. E-mail: uno@cens.tokai.jaeri.go.jp; Kaneko, J.; Nishitani, T.; Maekawa, F.; Tanaka, T.; Shibata, S.; Ikeda, Y.; Khripunov, V.; Walker, C.; Ebisawa, K.; Takeuchi, H

    2001-10-01

    Neutron activation with water flow based on the {sup 16}O(n, p){sup 16}N reaction (T{sub 1/2}=7.13 s) has been proposed for the accurate fusion power monitor with reasonable time resolution in International Thermonuclear Experimental Reactor (ITER)-FEAT. The experiment for validation of the method for determining the neutron flux was carried out at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Research Institute (JAERI). The stainless steel (SS316)/Water assembly was utilized to simulate a neutron field in the blanket region of ITER. The closed water loop was laid between the assembly and the shielded measurement area. The {gamma}-rays from the irradiated water were measured by Bi{sub 4}Ge{sub 3}O{sub 12} (BGO) scintillator which was {approx}12 m away from the irradiation point. The range of the flow velocity was 2-11 m/s. The neutron spectrum at the probe section in the SS316/Water assembly was calculated by the MCNP-4B code with the FENDL/E-2.0 library. The cross section data of the {sup 16}O(n, p){sup 16}N reaction was taken from the FENDL/A-2.0 file. The counting response of the BGO detector including the effect of the shielding lead blocks surrounding it was calculated by MCNP-4B. The neutron yield from the water activation has a good agreement with that from associated {alpha} particle detector.

  20. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  1. Analysis of carbon dioxide, water vapour and energy fluxes over an ...

    Indian Academy of Sciences (India)

    Analysis of carbon dioxide, water vapour and energy fluxes over an Indian teak mixed deciduous forest for winter and summer months using eddy covariance technique ... Eddy covariance; Indian deciduous forests; CO2 flux; heat flux. ... National Remote Sensing Centre (ISRO), Balanagar, Hyderabad 500 037, India.

  2. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Science.gov (United States)

    2013-06-14

    ... COMMISSION All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos... Licensees operate boiling-water reactors (BWRs) with Mark I and Mark II containment designs. II. The events... Boiling Water Reactors with Mark I and Mark II Containments'' (November 26, 2012). Option 2 in SECY-12...

  3. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff...)-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors.'' DG-1277...

  4. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power reactors...

  5. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide...

  6. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  7. The New Water Moderator of the IBR-2 Reactor with a Canyon on the Lateral Surface. Design and Physical Parameters

    CERN Document Server

    Korneev, D A; Bodnarchuk, V I; Peresedov, V F; Rogov, A D; Shabalin, E P; Yaradaikin, S P

    2003-01-01

    An element of the new cold methane moderator of the reactor IBR-2, the water premoderator, serves as a thermal moderator for the 9th and 1st channels. Neutron radiation in the direction of the 9th channel comes from the lateral surface of the moderator. A specific feature of the reflectometer REFLEX located on the 9th channel is that it only "sees" neutrons emitted from a limited region of the moderator surface. This region is a rectangular extended along a vertical with a horizontal dimension of about 7 mm. To increase the flux on the sample, a groove-like pocket (canyon) with a depth of 80 mm by the width 15 mm and height 200 mm was cut in the premoderator on its lateral surface. The design of the moderator and the results of measurements of the neutron flux distribution on the lateral surface of the moderator are presented.

  8. Summary of the 3rd workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  9. KüFA safety testing of HTR fuel pebbles irradiated in the High Flux Reactor in Petten

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O., E-mail: oliver.seeger@rwth-aachen.de [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Abjani, A. El; Ejton, J.; Boudaud, D.; Freis, D.; Carbol, P.; Rondinella, V.V. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Fütterer, M. [European Commission, Joint Research Centre (JRC), Institute for Energy and Transport (IET), Nuclear Reactor Integrity Assessment and Knowledge Management Unit, PO Box 2, 1755 ZG Petten (Netherlands); Allelein, H.-J. [Lehrstuhl für Reaktorsicherheit und -technik an der RWTH Aachen, Kackertstraße 9, 52072 Aachen (Germany)

    2016-09-15

    The Cold Finger Apparatus (KühlFinger-Apparatur—KüFA) in operation at JRC-ITU is designed to experimentally scrutinize the effects of Depressurization LOss of Forced Circulation (D-LOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. Up to 1600 °C, the reference maximum temperature for these accidents, high-quality German HTR fuel pebbles have already demonstrated a small fission product release. This paper discusses and compares the releases obtained from KüFA-testing the pebbles HFR-K5/3 and HFR-EU1/3, which were both irradiated in the High Flux Reactor (HFR) in Petten. We present the time-dependent fractional release of the volatile fission product {sup 137}Cs as well as the fission gas {sup 85}Kr for both pebbles. For HFR-EU1/3 the isotopes {sup 134}Cs and {sup 154}Eu as well as the shorter-lived {sup 110m}Ag have also been measured. A detailed description of the experimental setup and its accuracy is given. The data for the recently tested pebbles is discussed in the context of previous results.

  10. Multi-Application Small Light Water Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept

  11. Comparison of ecosystem water flux measured with the Eddy covariance- and the direct xylem sap flux method in a mountainous forest

    Energy Technology Data Exchange (ETDEWEB)

    Stefanicki, G.; Geissbuehler, P.; Siegwolf, R. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1999-08-01

    The Eddy covariance technique allows to measure different components of turbulent air fluxes, including the flow of water vapour. Sap flux measurements determine directly the water flow in tree stems. We compared the water flux just above the crowns of trees in a forest by the technique of Eddy covariance and the water flux by the xylem sap flux method. These two completely different approaches showed a good qualitative correspondence. The correlation coefficient is 0.8. With an estimation of the crown diameter of the measured tree we also find a very good quantitative agreement. (author) 3 figs., 5 refs.

  12. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  13. Flux

    DEFF Research Database (Denmark)

    Ravn, Ib

    Med udgangspunkt i kompleksistetsforskning og studiet af selvorganiserende systemer beskriver lb Ravn den fysiske og biologiske evolution og menneskets udvikling. Han fortolker begreber som kultur, sprog, frihed, værdier, mening, smerte og det ondes problem i lyset af en procesbaseret ontologi...... kanalisering af den flux eller energi, der strømmer igennem os og giver sig til kende i vore daglige aktiviteter. Skal vores tanker, handlinger, arbejde, samvær og politiske liv organiseres efter stramme og faste regelsæt, uden slinger i valsen? Eller skal de tværtimod forløbe ganske uhindret af regler og bånd...

  14. Neutron Flux Monitoring Based on Blind Source Separation Algorithms in Moroccan TRIGA MARK II Reactor

    Directory of Open Access Journals (Sweden)

    Hanane Arahmane

    2017-01-01

    Full Text Available We present an overview of fission chamber’s functioning modes, theoretical aspects of the nonnegative matrix factorization methods, and the opportunities that offer neutron data processing in order to achieve neutron flux monitoring tasks. Indeed, it is a part of research project that aimed at applying Blind Source Separation methods for in-core and ex-core neutron flux monitoring while analyzing the outputs of fission chamber. The latter could be used as a key issue for control, fuel management, safety concerns, and material irradiation experiments. The Blind Source Separation methods had been used in many scientific fields such as biomedical engineering and telecommunications. Recently, they were used for gamma spectrometry data processing. The originality of this research work is to apply these powerful methods to process the fission chamber output signals. We illustrated the effectiveness of this tool using simulated fission chamber signals.

  15. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  16. Assessment of similarity of HFBR (High Flux Beam Reactor) with separate effects test

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Slovik, G.C.

    1990-11-01

    A Separate Effects Test (SET) facility was constructed in 1963 to demonstrate the feasibility of the HFBR design and to determine the core power limits for a safe flow reversal event. The objective of the task reported here is to review the capability of the test to scale the dominant phenomena in the HFBR during a flow reversal event and the applicability of the range of the power level obtained from the test to the HFBR. The conclusion of this report was that the flow during the flow reversal event will not be similar in the two facilities. The causes of the dissimilarity are the differences in the core inlet friction, bypass path friction, the absence of the check valve in the test, and the materials used to represent the fuel plates. The impact of these differences is that the HFBR will undergo flow reversal sooner than the test and will have a higher flow rate in the final Natural Circulation Period. The shorter duration of the flow reversal event will allow less time for the plate to heat up and the larger flow in the Natural Circulation Period will lead to higher critical heat flux limits in the HFBR than in the test. Based on these observations, it was concluded that the HFBR can undergo flow reversal safely for heat fluxes up to 46,700 (BTU/hr ft{sup 2}), the heat flux limit obtained from the 1963 test.

  17. Bacterial Colonization of Pellet Softening Reactors Used during Drinking Water Treatment

    NARCIS (Netherlands)

    Hammes, F.; Boon, N.; Vital, M.; Ross, P.; Magic-Knezev, A.; Dignum, M.

    2010-01-01

    Pellet softening reactors are used in centralized and decentralized drinking water treatment plants for the removal of calcium (hardness) through chemically induced precipitation of calcite. This is accomplished in fluidized pellet reactors, where a strong base is added to the influent to increase

  18. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-04-19

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... Management Criteria for PWR Reactor Vessel Internal Components.'' The original notice provided the ADAMS... published a notice requesting public comments on draft LR-ISG-2011-04, ``Updated Aging Management Criteria...

  19. Supercritical water gasification of sewage sludge in continuous reactor.

    Science.gov (United States)

    Amrullah, Apip; Matsumura, Yukihiko

    2017-10-05

    In this study, a process for the continuous recovery of phosphorus and generation of gas from sewage sludge is investigated for the first time using supercritical water gasification (SCWG). A continuous reactor was employed and experiments were conducted by varying the temperature (500-600 °C) and residence time (5-60 s) while fixing the pressure at 25 MPa. The behavior of phosphorus during the SCWG process was studied. The effect of the temperature and time on the composition of the product gas was also investigated. A model of the reaction kinetics for the SCWG of sewage sludge was developed. The organic phosphorus (OP) was rapidly converted into inorganic phosphorus (IP) within a short residence time of 10 s. The gaseous products were mainly composed of H2, CO2, and CH4. The reaction followed first order kinetics, and the model was found to fit the experimental data well. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. TOBI: A nonlinear model for boiling water reactor stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wehle, F. (Siemens/KWU, Offenbach (West Germany)); Pruitt, D.W.

    1990-06-01

    The magnitude and the divergent nature of the oscillations during the LaSalle unit 2 nuclear power plant even on March 9, 1988, renewed concern about the state of knowledge on boiling water reactor (BWR) instabilities and was followed by many activities, e.g., the Idaho Stability Symposium. For appropriate representation of the physical processes, typical BWR time-domain stability calculations with, e.g., TRAC, RETRAN, or THERMIT require a large number of axial nodes and are very costly with regard to computer time. Linear models are inexpensive, but only valid as long as the parameters have no large deviation from the reference operating conditions. The objective of this work is the development of a physical model that is applicable for stability analysis in the nonlinear regime, but without the disadvantage of numerical problems and excessive computing times. The basic concept of the model TOBI is the integral study of the interaction between the time-dependent single- and two-phase regions. A series of calculations for purely thermal-hydraulic systems and BWRs has shown that TOBI is a convenient tool for stability analysis. Because of the simple but physically realistic modeling, it is very helpful in achieving an improved understanding of the mechanisms that affect BWR stability, and it is inexpensive, with regard to computer time, to perform extensive parameter sensitivity studies, even in the nonlinear regime.

  1. Structural analysis of fuel rod applied to pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Danilo P.; Pinheiro, Andre Ricardo M.; Lotto, André A., E-mail: danilo.pinheiro@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The design of fuel assemblies applied to Pressurized Water Reactors (PWR) has several requirements and acceptance criteria that must be attended for licensing. In the case of PWR fuel rods, an important mechanical structural requirement is to keep the radial stability when submitted to the coolant external pressure. In the framework of the Accident Tolerant Fuel (ATF) program new materials have been studied to replace zirconium based alloys as cladding, including iron-based alloys. In this sense, efforts have been made to evaluate the behavior of these materials under PWR conditions. The present work aims to evaluate the collapse cold pressure of a stainless steel thin-walled tube similar to that used as cladding material of fuel rods by means of the comparison of numeric data, and experimental results. As a result of the simulations, it was observed that the collapse pressure has a value intermediate value between those found by regulatory requirements and analytical calculations. The experiment was carried out for the validation of the computational model using test specimens of thin-walled tubes considering empty tube. The test specimens were sealed at both ends by means of welding. They were subjected to a high pressure device until the collapse of the tubes. Preliminary results obtained from experiments with the empty test specimens indicate that the computational model can be validated for stainless steel cladding, considering the difference between collapse pressure indicated in the regulatory document and the actual limit pressure concerning to radial instability of tubes with the studied characteristics. (author)

  2. Study on unstable fracture characteristics of light water reactor piping

    Energy Technology Data Exchange (ETDEWEB)

    Kurihara, Ryoichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-08-01

    Many testing studies have been conducted to validate the applicability of the leak before break (LBB) concept for the light water reactor piping in the world. It is especially important among them to clarify the condition that an inside surface crack of the piping wall does not cause an unstable fracture but ends in a stable fracture propagating only in the pipe thickness direction, even if the excessive loading works to the pipe. Pipe unstable fracture tests performed in Japan Atomic Energy Research Institute had been planned under such background, and clarified the condition for the cracked pipe to cause the unstable fracture under monotonous increase loading or cyclic loading by using test pipes with the inside circumferential surface crack. This paper examines the pipe unstable fracture by dividing it into two parts. One is the static unstable fracture that breaks the pipe with the inside circumferential surface crack by increasing load monotonously. Another is the dynamic unstable fracture that breaks the pipe by the cyclic loading. (author). 79 refs.

  3. CFD simulation on critical heat flux of flow boiling in IVR-ERVC of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xiang, E-mail: zhangxiang3@snptc.com.cn [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Hu, Teng [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Chen, Deqi, E-mail: chendeqi@cqu.edu.cn [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China); Zhong, Yunke; Gao, Hong [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China)

    2016-08-01

    Highlights: • CFD simulation on CHF of boiling two-phase flow in ERVC is proposed. • CFD simulation result of CHF agrees well with that of experimental result. • The characteristics of boiling two-phase flow and boiling crisis are analyzed. - Abstract: The effectiveness of in-vessel retention (IVR) by external reactor vessel cooling (ERVC) strongly depends on the critical heat flux (CHF). As long as the local CHF does not exceed the local heat flux, the lower head of the pressure vessel can be cooled sufficiently to prevent from failure. In this paper, a CFD simulation is carried out to investigate the CHF of ERVC. This simulation is performed by a CFD code fluent couple with a boiling model by UDF (User-Defined Function). The experimental CHF of ERVC obtained by State Nuclear Power Technology Research and Development Center (SNPTRD) is used to validate this CFD simulation, and it is found that the simulation result agrees well with the experimental result. Based on the CFD simulation, detailed analysis focusing on the pressure distribution, velocity distribution, void fraction distribution, heating wall temperature distribution are proposed in this paper.

  4. Adaptation of a High Frequency Ultrasonic Transducer to the Measurement of Water Temperature in a Nuclear Reactor

    Science.gov (United States)

    Zaz, G.; Calzavara, Y.; Le Clézio, E.; Despaux, G.

    Most high flux reactors possess for research purposes fuel elements composed of plates. Their relative distance is a crucial parameter, particularly concerning the irradiation history. For the High Flux Reactor (RHF) of the Institute Laue-Langevin (ILL), the measurement of this distance with a microscopic resolution becomes extremely challenging. To address this issue, a specific ultrasonic transducer, presented in a first paper, has been designed and manufactured to be inserted into the 1.8 mm width channel existing between curved fuel plates. It was set on a blade yielding a total device thickness of 1 mm. To achieve the expected resolution, the system is excited with frequencies up to 70 MHz and integrated into a set of high frequency acquisition instruments. Thanks to a specific signal processing, this device allows the distance measurement through the evaluation of the ultrasonic wave time of fight. One of the crucial points is then the evaluation of the local water temperature inside the water channel. To obtain a precise estimation of this parameter, the ultrasonic sensor is used as a thermometer thanks to the analysis of the spectral components of the acoustic signal propagating inside the sensor multilayered structure. The feasibility of distance measurement was proved during the December 2013 experiment in the RHF fuel element of the ILL. Some of the results will be presented as well as some experimental constraints identified to improve the accuracy of the measurement in future works.

  5. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  6. Effect of spatial sampling from European flux towers for estimating carbon and water fluxes with artificial neural networks

    NARCIS (Netherlands)

    Papale, Dario; Black, T.A.; Carvalhais, Nuno; Cescatti, Alessandro; Chen, Jiquan; Jung, Martin; Kiely, Gerard; Lasslop, Gitta; Mahecha, Miguel D.; Margolis, Hank; Merbold, Lutz; Montagnani, Leonardo; Moors, Eddy; Olesen, J.E.; Reichstein, Markus; Tramontana, Gianluca; Gorsel, Van Eva; Wohlfahrt, Georg; Ráduly, Botond

    2015-01-01

    Empirical modeling approaches are frequently used to upscale local eddy covariance observations of carbon, water, and energy fluxes to regional and global scales. The predictive capacity of such models largely depends on the data used for parameterization and identification of input-output

  7. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  8. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  9. Interactive effects of elevated CO2 and drought on nocturnal water fluxes in Eucalyptus saligna.

    Science.gov (United States)

    Zeppel, Melanie J B; Lewis, James D; Medlyn, Belinda; Barton, Craig V M; Duursma, Remko A; Eamus, Derek; Adams, Mark A; Phillips, Nathan; Ellsworth, David S; Forster, Michael A; Tissue, David T

    2011-09-01

    Nocturnal water flux has been observed in trees under a variety of environmental conditions and can be a significant contributor to diel canopy water flux. Elevated atmospheric CO(2) (elevated [CO(2)]) can have an important effect on day-time plant water fluxes, but it is not known whether it also affects nocturnal water fluxes. We examined the effects of elevated [CO(2)] on nocturnal water flux of field-grown Eucalyptus saligna trees using sap flux through the tree stem expressed on a sapwood area (J(s)) and leaf area (E(t)) basis. After 19 months growth under well-watered conditions, drought was imposed by withholding water for 5 months in the summer, ending with a rain event that restored soil moisture. Reductions in J(s) and E(t) were observed during the severe drought period in the dry treatment under elevated [CO(2)], but not during moderate- and post-drought periods. Elevated [CO(2)] affected night-time sap flux density which included the stem recharge period, called 'total night flux' (19:00 to 05:00, J(s,r)), but not during the post-recharge period, which primarily consisted of canopy transpiration (23:00 to 05:00, J(s,c)). Elevated [CO(2)] wet (EW) trees exhibited higher J(s,r) than ambient [CO(2)] wet trees (AW) indicating greater water flux in elevated [CO(2)] under well-watered conditions. However, under drought conditions, elevated [CO(2)] dry (ED) trees exhibited significantly lower J(s,r) than ambient [CO(2)] dry trees (AD), indicating less water flux during stem recharge under elevated [CO(2)]. J(s,c) did not differ between ambient and elevated [CO(2)]. Vapour pressure deficit (D) was clearly the major influence on night-time sap flux. D was positively correlated with J(s,r) and had its greatest impact on J(s,r) at high D in ambient [CO(2)]. Our results suggest that elevated [CO(2)] may reduce night-time water flux in E. saligna when soil water content is low and D is high. While elevated [CO(2)] affected J(s,r), it did not affect day-time water

  10. Carbon and Water Vapor Fluxes of Different Ecosystems in Oklahoma

    Science.gov (United States)

    Wagle, P.; Gowda, P. H.; Northup, B. K.

    2016-12-01

    Information on exchange of energy, carbon dioxide (CO2), and water vapor (H2O) for major terrestrial ecosystems is vital to quantify carbon and water balances on a large-scale. It is also necessary to develop, test, and improve crop models and satellite-based production efficiency and evapotranspiration (ET) models, and to better understand the potential of terrestrial ecosystems to mitigate rising atmospheric CO2 concentration and climate change. A network (GRL-FLUXNET) of nine eddy flux towers has been established over a diverse range of terrestrial ecosystems, including native and improved perennial grasslands [unburned and grazed tallgrass prairie, burned and grazed tallgrass prairie, and burned Bermuda grass (Cynodon dactylon L.)], grazed and non-grazed winter wheat (Triticum aestivum L.), till and no-till winter wheat and canola (Brassica napus L.), alfalfa (Medicago sativa L.), and soybean (Glycine max L.), at the USDA-ARS, Grazinglands Research Laboratory, El Reno, OK. In this presentation, we quantify and compare net ecosystem CO2 exchange (NEE) and ET between recently burned and grazed tallgrass prairie and burned and non-grazed Bermuda grass pastures, alfalfa, and soybean. Preliminary results show monthly ensembles average NEE reached seasonal peak values of -29, -35, -25, and -20 µmol m-2 s-1 in burned tallgrass prairie pasture, burned Bermuda grass pasture, alfalfa, and soybean, respectively. Similarly, monthly ensembles average ET reached seasonal peak values of 0.22, 0.27, 0.25, 0.28 mm 30-min-1 in burned tallgrass prairie pasture, burned Bermuda grass pasture, alfalfa, and soybean, respectively. Seasonal patterns and daily magnitudes of NEE and ET and their responses to the similar climatic conditions will be further investigated.

  11. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  12. Indirect influence of eutrophication on air - water exchange fluxes, sinking fluxes, and occurrence of polycyclic aromatic hydrocarbons.

    Science.gov (United States)

    Tao, Yuqiang; Yu, Jing; Lei, Guoliang; Xue, Bin; Zhang, Fengju; Yao, Shuchun

    2017-10-01

    How eutrophication affects biogeochemical processes of hydrophobic organic contaminants (HOCs) in aquatic environments is a pending challenge. Although the direct influence of eutrophication on biogeochemical processes of HOCs in waters has been well addressed, the indirect influence of eutrophication on biogeochemical processes of HOCs remains largely unknown. Here we take the large shallow eutrophic Lake Taihu in China and polycyclic aromatic hydrocarbons (PAHs) as examples to provide novel knowledge on the indirect influence of eutrophication on air - water exchange fluxes, sinking fluxes, and occurrence of HOCs. The air - water exchange fluxes of individual PAHs varied dramatically at different sites in all studied seasons. The sinking flux of ΣPAH 16 was 14 855.3 ± 1579.9, 3548.9 ± 650.6, and 5588.4 ± 530.7 ng m -2  d -1 in spring, summer, and winter. The corresponding concentration of ΣPAH 16 in surface sediments was 713.1 ± 78.6, 339.7 ± 36.6, and 293.0 ± 35.2 ng g -1  d.w. Our study for the first time suggested that recruitment of cyanobacteria from surface sediments to water column in spring reduced the concentrations of PAHs in surface sediments, but enhanced their concentrations in the bulk water column, and overwintering of cyanobacteria in winter enhanced the concentrations of PAHs in surface sediments. High pH induced indirectly by eutrophication decreased seasonal air - water exchange fluxes (enhanced net volatilization) of PAHs, reduced the aromaticity of surface sediments and the hydrophobicity of phytoplankton cell surface, and reduced the accumulation of PAHs in surface sediments and phytoplankton consequently. Sinking fluxes and daily loss of PAHs from the water column decreased with phytoplankton biomass because the fraction of organic matter sinking from the water column decreased with phytoplankton biomass. Our study provides novel complementary knowledge for the biological pump for HOCs, and has important

  13. Summary of the 4th workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  14. Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor; Diseno y construccion de un sistema electronico automatico de medicion y graficado del flujo neutronico para el reactor subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1997-07-01

    The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

  15. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    Energy Technology Data Exchange (ETDEWEB)

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  16. Pressurized hydrogenotrophic denitrification reactor for small water systems.

    Science.gov (United States)

    Epsztein, Razi; Beliavski, Michael; Tarre, Sheldon; Green, Michal

    2017-03-15

    The implementation of hydrogenotrophic denitrification is limited due to safety concerns, poor H2 utilization and low solubility of H2 gas with the resulting low transfer rate. The current paper presents the main research work conducted on a pressurized hydrogenotrophic reactor for denitrification that was recently developed. The reactor is based on a new concept suggesting that a gas-liquid equilibrium is achieved in the closed headspace of denitrifying reactor, further produced N2 gas is carried out by the effluent and gas purging is not required. The feasibility of the proposed reactor was shown for two effluent concentrations of 10 and 1 mg NO3--N/L. Hydrogen gas utilization efficiencies of 92.8% and 96.9% were measured for the two effluent concentrations, respectively. Reactor modeling predicted high denitrification rates above 4 g NO3--N/(Lreactor·d) at reasonable operational conditions. Hydrogen utilization efficiency was improved up to almost 100% by combining the pressurized reactor with a following open-to-atmosphere polishing unit. Also, the potential of the reactor to remove ClO4- was shown. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

    2008-03-15

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

  18. Measurements and calculations of water velocity, momentum flux, and related flow parameters obtaned from single-phase water integral acceptance tests of the PKL instrumented spool pieces

    Energy Technology Data Exchange (ETDEWEB)

    Stein, W.

    1979-10-09

    The operation of the emergency core cooling system and its related steam-binding problems in pressurized water reactors is the subject of a cooperative study by the United States, Germany, and Japan. Lawrence Livermore Laboratory and EG and G, Inc., San Ramon Operations, are responsible for the design, hardware, and software of the 80.8-mm and 113-mm spool piece measurement systems for the German Primarkreislauf (PKL) Test Facility at Kraftwerk Union in Erlangen, West Germany. This work was done for the US Nuclear Regulatory Commission, Division of Reactor Safety Research, under its 3-D Technical Support and Instrumentation Program. Four instrumented spools capable of measuring individual phase parameters in two-phase flows were constructed. Each spool contains a flow turbine, drag screen, three-beam densitometer, and pressure and temperature probes. A computerized data acquisition system is also provided to store and analyze data from the four spools. The four spools were shipped to the PKL Test Facility in West Germany for acceptance testing in a water-flow loop. Spool measurements of velocity and momentum flux were compared to the values obtained from an orifice meter installed in the loop piping system. The turbine flowmeter velocity data for all tests were within allowable tolerances. Drag screen momentum flux measurements were also within tolerance with the exception of a few points.

  19. Pool boiling of distilled water over tube bundle with variable heat flux

    Science.gov (United States)

    Swain, Abhilas; Mohanty, Rajiva Lochan; Das, Mihir Kumar

    2017-08-01

    The experimental investigation of saturated pool boiling heat transfer of distilled water over plain tube bundle, under uniform and varying heat flux condition along the height are presented in this article. Experiments are carried out under various heat flux configurations applied to rows of tube bundles and pitch distance to diameter ratios of 1.25, 1.6 and 1.95. The wall superheats and pool boiling heat transfer coefficients over individual rows are determined. The pool boiling heat transfer coefficients for variable heat flux and uniform heat flux conditions are compared. The results indicate that the bundle effect is found to exist for uniform as well as variable heat flux under all operating conditions in the present investigation. The variable heat flux resulted in range of wall superheat being highest for decreasing heat flux from bottom to top and lowest for increasing heat flux from bottom to top.

  20. Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

    Science.gov (United States)

    Lizon-A-Lugrin, Laure

    The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

  1. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  2. Thermal hydraulic test for reactor safety system - Critical heat flux experiment and development of prediction models

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Baek, Won Pil; Yang, Soo Hyung; No, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    2000-04-01

    To acquire CHF data through the experiments and develop prediction models, research was conducted. Final objectives of research are as follows: 1) Production of tube CHF data for low and middle pressure and mass flux and Flow Boiling Visualization. 2) Modification and suggestion of tube CHF prediction models. 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. The major results of research are as follows: 1) Production of the CHF data for low and middle pressure and mass flux. - Acquisition of CHF data (764) for low and middle pressure and flow conditions - Analysis of CHF trends based on the CHF data - Assessment of existing CHF prediction methods with the CHF data 2) Modification and suggestion of tube CHF prediction models. - Development of a unified CHF model applicable for a wide parametric range - Development of a threshold length correlation - Improvement of CHF look-up table using the threshold length correlation 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. - Development of bundle CHF prediction methodology using correction factor. 11 refs., 134 figs., 25 tabs. (Author)

  3. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  4. Effect of spatial sampling from European flux towers for estimating carbon and water fluxes with artificial neural network

    DEFF Research Database (Denmark)

    Papale, D.; Black, T Andrew; Carvalhais, Nuno

    2015-01-01

    Empirical modeling approaches are frequently used to upscale local eddy covariance observations of carbon, water, and energy fluxes to regional and global scales. The predictive capacity of such models largely depends on the data used for parameterization and identification of input......-output relationships, while prediction for conditions outside the training domain is generally uncertain. In this work, artificial neural networks (ANNs) were used for the prediction of gross primary production (GPP) and latent heat flux (LE) on local and European scales with the aim to assess the portion...

  5. Material gap membrane distillation: A new design for water vapor flux enhancement

    KAUST Repository

    Francis, Lijo

    2013-08-19

    A new module design for membrane distillation, namely material gap membrane distillation (MGMD), for seawater desalination has been proposed and successfully tested. It has been observed that employing appropriate materials between the membrane and the condensation plate in an air gap membrane distillation (AGMD) module enhanced the water vapor flux significantly. An increase in the water vapor flux of about 200-800% was observed by filling the gap with sand and DI water at various feed water temperatures. However, insulating materials such as polypropylene and polyurethane have no effect on the water vapor flux. The influence of material thickness and characteristics has also been investigated in this study. An increase in the water gap width from 9. mm to 13. mm increases the water vapor flux. An investigation on an AGMD and MGMD performance comparison, carried out using two different commercial membranes provided by different manufacturers, is also reported in this paper. © 2013 Elsevier B.V.

  6. Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ; Caracterizacion del flujo neutronico en el Hohlraum de la columna termica del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Palacios, J.C.; Alonso, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

    2006-07-01

    Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

  7. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  8. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  9. Recent results of research on supercritical water-cooled reactors in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Koehly, C.; Schulenberg, T. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Maraczy, C. [AEKI-KFKI, Budapest (Hungary); Toivonen, A.; Penttila, S. [VTT Technical Research Centre, Espoo (Finland); Chandra, L.; Lycklama a Nijeholt, J.A. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands)

    2009-07-01

    In Europe, the research on Supercritical Water-Cooled Reactors is integrated in a project called 'High Performance Light Water Reactor Phase 2' (HPLWR Phase 2), co-funded by the European Commission. Ten partners and three active supporters are working on critical scientific issues to determine the potential of this reactor concept in the electricity market. The recent design of the HPLWR including flow paths is described in this paper. Exemplarily, design analyses are presented addressing neutronics, thermal-hydraulics, thermo-mechanics, materials investigations and heat transfer. (author)

  10. Analysis of neutron flux distribution using the Monte Carlo method for the feasibility study of the Prompt Gamma Activation Analysis technique at the IPR-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guerra, Bruno T.; Pereira, Claubia, E-mail: brunoteixeiraguerra@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Departmento de Energia Nuclear; Soares, Alexandre L.; Menezes, Maria Angela B.C., E-mail: menezes@cdtn.br, E-mail: asleal@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor and operates at 100 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in {sup 235}U. The implementation of the PGNAA (Prompt Gamma Neutron Activation Analysis) using this research reactor will significantly increase in number of chemical elements analysed and the kind of matrices. A project is underway in order to implement this technique at CDTN. The objective of this study was to contribute in feasibility analysis of implementing this technique. For this purpose, MCNP is being used. Some variance reduction tools in the methodology, that has been already developed, was introduced for calculating of the neutron flux in the neutron extractor inclined. The objective was to reduce the code error and thereby increasing the reliability of the results. With the implementation of the variance reduction tools, the results of the thermal and epithermal neutron fluxes presented a significant improvement in both calculations. (author)

  11. Numerical Simulation of Water Jet Flow Using Diffusion Flux Mixture Model

    Directory of Open Access Journals (Sweden)

    Zhi Shang

    2014-01-01

    Full Text Available A multidimensional diffusion flux mixture model was developed to simulate water jet two-phase flows. Through the modification of the gravity using the gradients of the mixture velocity, the centrifugal force on the water droplets was able to be considered. The slip velocities between the continuous phase (gas and the dispersed phase (water droplets were able to be calculated through multidimensional diffusion flux velocities based on the modified multidimensional drift flux model. Through the numerical simulations, comparing with the experiments and the simulations of traditional algebraic slip mixture model on the water mist spray, the model was validated.

  12. Passive gamma analysis of the boiling-water-reactor assemblies

    Science.gov (United States)

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  13. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1976

    Energy Technology Data Exchange (ETDEWEB)

    Ferguson, J. B. [ed.

    1977-04-01

    Light water reactor safety research performed October through December 1976 is discussed. An analysis to determine the effect of emergency core coolant (ECC) injection location and pump speed on system response characteristics was performed. An analysis to evaluate the capability of commonly used critical heat flux (CHF) correlations to calculate the time of the first CHF in the Semiscale core during a loss-of-coolant experiment (LOCE) was performed. A test program and study to determine the effect thermocouples mounted on the outside fuel rod surfaces would have on the departure from nucleate boiling (DNB) phenomena in the LOFT core during steady state operation were completed. A correlation for use in predicting DNB heat fluxes in the LOFT core was developed. Tests of an experimental transit time flowmeter were completed. A nuclear test was performed to obtain fuel rod behavior data from four PWR-type rods during film boiling operation representative of PWR conditions. Preliminary results from the postirradiation examination of Test IE-1 fuel rods are given. Results of Irradiation Effects Tests IE-2 and IE-3 are given. Gap Conductance Test GC 2-1 was performed to evaluate the effects of fuel density, initial gap width, and fill gas composition on the pellet-cladding gap conductance.

  14. Visualizing ocular lens fluid dynamics using MRI: manipulation of steady state water content and water fluxes.

    Science.gov (United States)

    Vaghefi, Ehsan; Pontre, Beau P; Jacobs, Marc D; Donaldson, Paul J

    2011-08-01

    Studies using various MRI techniques have shown that a water-protein concentration gradient exists in the ocular lens. Because this concentration is higher in the core relative to the lens periphery, a gradient in refractive index is established in the lens. To investigate how the water-protein concentration profile is maintained, bovine lenses were incubated in different solutions, and changes in water-protein concentration ratio monitored using proton density weighted (PD-weighted) imaging in the absence and presence of heavy water (D(2)O). Lenses incubated in artificial aqueous humor (AAH) maintained the steady state water-protein concentration gradient, but incubating lenses in high extracellular potassium (KCl-AAH) or low temperature (Low T-AAH) caused a collapse of the gradient due to a rise in water content in the core of the lens. To visualize water fluxes, lenses were incubated in D(2)O, which acts as a contrast agent. Incubation in KCl-AAH and low T-AAH dramatically slowed the movement of D(2)O into the core but did not affect the movement of D(2)O into the outer cortex. D(2)O seemed to preferentially enter the lens cortex at the anterior and posterior poles before moving circumferentially toward the equatorial regions. This directionality of D(2)O influx into the lens cortex was abolished by incubating lenses in high KCl-AAH or low T-AAH, and resulted in homogenous influx of D(2)O into the outer cortex. Taken together, our results show that the water-protein concentration ratio is actively maintained in the core of the lens and that water fluxes preferentially enter the lens at the poles.

  15. Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Calzavara, Y. [Inst. Laue-Langevin (ILL), Grenoble (France)

    2014-09-30

    The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied to the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.

  16. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; 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 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  17. Critical heat flux experiments in a circular tube with heavy water and light water. (AWBA Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, C.L.; Beus, S.G.

    1980-05-01

    Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on the average, 8 percent below the light water CHF values.

  18. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  19. Climatological evaluation of some fluxes of the surface energy and soil water balances over France

    Directory of Open Access Journals (Sweden)

    E. M. Choisnel

    Full Text Available This paper presents some statistical evaluations of the surface energy and soil water balance fluxes, for a prairie-type canopy, using the Earth model with a double-reservoir system for the management of the soil water reserve and the regulation of actual evapotranspiration. The mean values of these fluxes are estimated from energy and water balance simulations done on a 30-year climatic reference period (1951–1980. From values of these fluxes calculated for each meteorological synoptic station, mappings of net radiation, actual evapotranspiration, drainage and conduction fluxes have been made over French territory. Lastly, a few conclusions pertaining to the spatial variability of fluxes and to the partition of rainfall between run-off and drainage on the one hand and replenishment of the soil water reserve on the other hand are drawn from these preliminary results.

  20. The role of water chemistry for environmentally assisted cracking in low-alloy reactor pressure vessel and piping steels under boiling reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S

    2005-07-01

    The environmentally assisted initiation and propagation of cracks in structural materials is one of the most important degradation and ageing mechanisms in light water reactors (LWR) and may seriously affect plant availability and economics. In the first part of this paper a short general introduction on environmentally assisted cracking (EAC) and its significance for LWR is given. Then the important role of water chemistry control in reducing the EAC risk in LWR is illustrated by current research results about the effect of chloride transients and hydrogen water chemistry on the EAC crack growth behaviour of low-alloy reactor pressure vessel and piping steels under boiling water reactor conditions. (author)

  1. Energy and Water Fluxes in Heterogeneous Mediterranean Water-limited Ecosystems

    Science.gov (United States)

    Detto, M.; Katul, G.; Mancini, M.

    2005-12-01

    Research efforts in distributed eco-hydrologic models often fall in one of two categories: prognostic, in which predictions of root-zone soil moisture content and land surface fluxes is required for a projected radiative and precipitation forcing time series, or diagnostic in which the relationship between soil water status and atmospheric water vapor demand is to be derived for the various components of the landscape. The latter relationships are now receiving broad attention in climate change, hydrological, and ecological studies of arid and semi-arid ecosystems. This interest is now a central focus given the recognition that the component latent heat flux sensitivity to soil moisture decline can directly impact plant productivity, carbon and nutrient cycling, and ground water recharge. With projected shifts in precipitation statistics, mainly towards increased desertification, the "stability" of these ecosystems is highly dependent on their ability to uptake water at low soil moisture Here, we determine the relationship between soil water status and atmospheric water vapor demand for patchy landscapes within a semi-arid ecosystems using a combination remote sensing products and field experiments. In particular, we investigate how VIS/NIR measurements, in conjunction with standard micrometeorological data and ground based thermal infrared thermometers, provide "diagnostic" hydrologic relationship between soil water content and potential evapo-transpiration for the various components of the landscape. These experiments were conducted in the Orroli site, situated in the mid-west of Sardinia (Italy) within the Flumendosa river watershed, which is considered one of the most important water supply resources to the island. The landscape is a mixture of Mediterranean patchy vegetation types: trees, including wild olives (/Olea sylvestris/) and cork oaks (/Quercus suber/), different shrubs (/Asparagus acutifolius, Rubus ulmifolius/) and herbaceous species (/Asphodelus

  2. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor; Determinacion de nitrogeno en harina de trigo mediante analisis por activacion empleando el flujo de neutrones rapidos de un reactor nuclear termico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, T

    1976-07-01

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  3. Integrated Water Gas Shift Membrane Reactors Utilizing Novel, Non Precious Metal Mixed Matrix Membrane

    Energy Technology Data Exchange (ETDEWEB)

    Ferraris, John P. [Univ. of Texas-Dallas, Richardson, TX (United States). Dept. of Chemistry

    2013-09-30

    Nanoparticles of zeolitic imidazolate frameworks and other related hybrid materials were prepared by modifying published synthesis procedures by introducing bases, changing stoichiometric ratios, or adjusting reaction conditions. These materials were stable at temperatures >300 °C and were compatible with the polymer matrices used to prepare mixed- matrix membranes (MMMs). MMMs tested at 300 °C exhibited a >30 fold increase in permeability, compared to those measured at 35 °C, while maintaining H2/CO2 selectivity. Measurements at high pressure (up to 30 atm) and high temperature (up to 300 °C) resulted in an increase in gas flux across the membrane with retention of selectivity. No variations in permeability were observed at high pressures at either 35 or 300 °C. CO2-induced plasticization was not observed for Matrimid®, VTEC, and PBI polymers or their MMMs at 30 atm and 300 °C. Membrane surface modification by cross-linking with ethanol diamine resulted in an increase in H2/CO2 selectivity at 35 °C. Spectrometric analysis showed that the cross-linking was effective to temperatures <150 °C. At higher temperatures, the cross-linked membranes exhibit a H2/CO2 selectivity similar to the uncross-linked polymer. Performance of the polybenzimidazole (PBI) hollow fibers prepared at Santa Fe Science and Technology (SFST, Inc.) showed increased flux and selectivity at 300 °C, which is comparable to a flat PBI membrane. A water-gas shift reactor has been built and currently being optimized for testing under DOE conditions.

  4. Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the cold source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and

  5. Design and installation of a hot water layer system at the Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila

    2013-01-01

    Full Text Available A hot water layer system (HWLS is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

  6. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  7. Development of high-fidelity multiphysics system for light water reactor analysis

    Science.gov (United States)

    Magedanz, Jeffrey W.

    There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining

  8. Analytical characterization of selective benthic flux components in estuarine and coastal waters

    Science.gov (United States)

    King, Jeffrey N.

    2011-01-01

    Benthic flux is the rate of flow across the bed of a water body, per unit area of bed. It is forced by component mechanisms, which interact. For example, pressure gradients across the bed, forced by tide, surface gravity waves, density gradients, bed–current interaction, turbulence, and terrestrial hydraulic gradients, drive an advective benthic flux of water and constituents between estuarine and coastal waters, and surficial aquifers. Other mechanisms also force benthic flux, such as chemical gradients, bioturbation, and dispersion. A suite of component mechanisms force a total benthic flux at any given location, where each member of the suite contributes a component benthic flux. Currently, the types and characteristics of component interactions are not fully understood. For example, components may interact linearly or nonlinearly, and the interaction may be constructive or destructive. Benthic flux is a surface water–groundwater interaction process. Its discharge component to a marine water body is referred to, in some literature, as submarine groundwater discharge. Benthic flux is important in characterizing water and constituent budgets of estuarine and coastal systems. Analytical models to characterize selective benthic flux components are reviewed. Specifically, these mechanisms are for the component associated with the groundwater tidal prism, and forced by surface gravity wave setup, surface gravity waves on a plane bed, and the terrestrial hydraulic gradient. Analytical models are applied to the Indian River Lagoon, Florida; Great South Bay, New York; and the South Atlantic Bight in South Carolina and portions of North Carolina.

  9. Using heat to characterize streambed water flux variability in four stream reaches.

    Science.gov (United States)

    Essaid, Hedeff I; Zamora, Celia M; McCarthy, Kathleen A; Vogel, Jason R; Wilson, John T

    2008-01-01

    Estimates of streambed water flux are needed for the interpretation of streambed chemistry and reactions. Continuous temperature and head monitoring in stream reaches within four agricultural watersheds (Leary Weber Ditch, IN; Maple Creek, NE; DR2 Drain, WA; and Merced River, CA) allowed heat to be used as a tracer to study the temporal and spatial variability of fluxes through the streambed. Synoptic methods (seepage meter and differential discharge measurements) were compared with estimates obtained by using heat as a tracer. Water flux was estimated by modeling one-dimensional vertical flow of water and heat using the model VS2DH. Flux was influenced by physical heterogeneity of the stream channel and temporal variability in stream and ground-water levels. During most of the study period (April-December 2004), flux was upward through the streambeds. At the IN, NE, and CA sites, high-stage events resulted in rapid reversal of flow direction inducing short-term surface-water flow into the streambed. During late summer at the IN site, regional ground-water levels dropped, leading to surface-water loss to ground water that resulted in drying of the ditch. Synoptic measurements of flux generally supported the model flux estimates. Water flow through the streambed was roughly an order of magnitude larger in the humid basins (IN and NE) than in the arid basins (WA and CA). Downward flux, in response to sudden high streamflows, and seasonal variability in flux was most pronounced in the humid basins and in high conductivity zones in the streambed.

  10. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  11. Land use effects on green water fluxes from agricultural production in Mato Grosso, Brazil

    Science.gov (United States)

    Lathuilliere, M. J.; Johnson, M. S.; Donner, S. D.

    2010-12-01

    The blue water/green water paradigm is increasingly used to differentiate between subsequent routing of precipitation once it reaches the soil. “Blue” water is that which infiltrates deep in the soil to become streams and aquifers, while “green” water is that which remains in the soil and is either evaporated (non-productive green water) or transpired by plants (productive green water). This differentiation in the fate of precipitation has provided a new way of thinking about water resources, especially in agriculture for which better use of productive green water may help to relieve stresses from irrigation (blue water). The state of Mato Grosso, Brazil, presents a unique case for the study of green water fluxes due to an expanding agricultural land base planted primarily to soybean, maize, sugar cane, and cotton. These products are highly dependent on green water resources in Mato Grosso where crops are almost entirely rain-fed. We estimate the change in green water fluxes from agricultural expansion for the 2000-2008 period in the state of Mato Grosso based on agricultural production data from the Instituto Brasileiro de Geografia e Estatísticas and a modified Penman-Monteith equation. Initial results for seven municipalities suggest an increase in agricultural green water fluxes, ranging from 1-10% per year, due primarily to increases in cropped areas. Further research is underway to elucidate the role of green water flux variations from land use practices on the regional water cycle.

  12. A continuous stirred hydrogen-based polyvinyl chloride membrane biofilm reactor for the treatment of nitrate contaminated drinking water.

    Science.gov (United States)

    Xia, Siqing; Zhang, YanHao; Zhong, FoHua

    2009-12-01

    A continuous stirred hydrogen-based polyvinyl chloride (PVC) membrane biofilm reactor (MBfR) was investigated to remove nitrate from the drinking water. The reactor was operated over 100 days, and the result showed that the average nitrate denitrification rate of 1.2 g NO(3)(-)-N/m(2) d and the total nitrogen (TN) removal of 95.1% were achieved with the influent nitrate concentration of 50 mg NO(3)(-)-N/L and the hydrogen pressure of 0.05 MPa. Under the same conditions, the average rate of hydrogen utilization by biofilm was 0.031 mg H(2)/cm(2) d, which was sufficient to remove 50 mg NO(3)(-)-N/L from the contaminated water with the effluent nitrate and nitrite concentrations below drinking water limit values. The average hydrogen utilization efficiency was achieved as high as 99.5%. Flux analysis demonstrated that, compared to sulfate reduction, nitrate reduction competed more strongly for hydrogen electron, and obtained more electrons in high influent nitrate loading.

  13. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  14. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  15. Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F.

    1984-07-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

  16. European research activities within the project: High Performance Light Water Reactor phase 2 (HPLWR phase 2)

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schulenberg, T. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, Karlsruhe (Germany); Marsault, P. [CEA Cadarache (DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Bittermann, D. [AREVA NP, NEPR-G, Erlangen (Germany); Maraczy, C. [AEKI-KFKI, Budapest (Hungary); Laurien, E. [Stuttgart Univ. IKE (Germany); Lycklama, J.A. [NRG Petten, NL (Netherlands); Anglart, H. [KTH Energy Technology, Stockholm (Sweden); Aksan, N. [Paul Scherrer Institut CH, Villigen PSI (Switzerland); Ruzickova, M. [UJV Rez plc, Husinec-Rez c.p. (Czech Republic); Heikinheimo, L. [VTT, FIN (Finland)

    2007-07-01

    The High Performance Light Water Reactor (HPLWR) is a Light Water Reactor (LWR) operating at supercritical pressure (25 MPa). It belongs to the six reactors currently being investigated under the framework of the Generation IV International Forum. The most visible advantage of the HPLWR shall be the low construction costs in the order of 1000 Euro/kWe, because of size reduction of components and buildings compared to current Light Water Reactors, and the low electricity production costs which are targeted at 3-4 cents/kWh. In Europe, investigations on the HPLWR have been integrated into a joint research project, called High Performance Light Water Reactor Phase 2 (HPLWR Phase 2), which is co-funded by the European Commission. Within 42 months, ten partners from eight European countries working on critical scientific issues shall show the feasibility of the HPLWR concept. This paper reports on 5 points relevant for HPLWR: 1) design and integration, 2) core design, 3) safety, 4) materials, and 5) heat transfer. The final goal is to assess the future potential of this reactor in the electricity market.

  17. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  18. EXPERIMENTAL STUDY OF CRITICAL HEAT FLUX WITH ALUMINA-WATER NANOFLUIDS IN DOWNWARD-FACING CHANNELS FOR IN-VESSEL RETENTION APPLICATIONS

    Directory of Open Access Journals (Sweden)

    G. DEWITT

    2013-06-01

    Full Text Available The Critical Heat Flux (CHF of water with dispersed alumina nanoparticles was measured for the geometry and flow conditions relevant to the In-Vessel Retention (IVR situation which can occur during core melting sequences in certain advanced Light Water Reactors (LWRs. CHF measurements were conducted in a flow boiling loop featuring a test section designed to be thermal-hydraulically similar to the vessel/insulation gap in the Westinghouse AP1000 plant. The effects of orientation angle, pressure, mass flux, fluid type, boiling time, surface material, and surface state were investigated. Results for water-based nanofluids with alumina nanoparticles (0.001% by volume on stainless steel surface indicate an average 70% CHF enhancement with a range of 17% to 108% depending on the specific flow conditions expected for IVR. Experiments also indicate that only about thirty minutes of boiling time (which drives nanoparticle deposition are needed to obtain substantial CHF enhancement with nanofluids.

  19. 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.

  20. Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

    Directory of Open Access Journals (Sweden)

    Hsoung-Wei Chou

    2015-01-01

    Full Text Available The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to be the severest accident for a boiling water reactor pressure vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic reactor pressure vessel has sufficient structural integrity until doubling of the present end-of-license operation.

  1. Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

    2004-11-01

    One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

  2. Forest floor methane flux modelled by soil water content and ground vegetation - comparison to above canopy flux

    Science.gov (United States)

    Halmeenmäki, Elisa; Peltola, Olli; Haikarainen, Iikka; Ryhti, Kira; Rannik, Üllar; Pihlatie, Mari

    2017-04-01

    Methane (CH4) is an important and strong greenhouse gas of which atmospheric concentration is rising. While boreal forests are considered as an important sink of CH4 due to soil CH4 oxidation, the soils have also a capacity to emit CH4. Moreover, vegetation is shown to contribute to the ecosystem-atmosphere CH4 flux, and it has been estimated to be the least well known natural sources of CH4. In addition to well-known CH4 emissions from wetland plants, even boreal trees have been discovered to emit CH4. At the SMEAR (Station for Measuring Ecosystem-Atmosphere Relations) II station in Hyytiälä, southern Finland (61° 51' N, 24°17' E; 181 m asl), we have detected small CH4 emissions from above the canopy of a Scots pine (Pinus sylvestris) dominated forest. To assess the origin of the observed emissions, we conducted forest floor CH4 flux measurements with 54 soil chambers at the footprint area of the above canopy flux measurements during two growing seasons. In addition, we measured the soil volumetric water content (VWC) every time next to the forest floor chamber measurements, and estimated vegetation coverages inside the chambers. In order to model the forest floor CH4 flux at the whole footprint area, we combined lidar (light detection and ranging) data with the field measurements. To predict the soil water content and thus the potential CH4 flux, we used local elevation, slope, and ground return intensity (GRI), calculated from the lidar data (National Land Survey of Finland). We categorized the soil chambers into four classes based on the VWC so that the class with the highest VWC values includes all the soil chambers with a potential to emit CH4. Based on a statistically significant correlation between the VWC and the forest floor CH4 flux (r = 0.30, p area. The results of the soil chamber measurements show a few areas of the forest floor with significant CH4 emissions. The modelled map of the potential CH4 flux is consistent with the measurements of the

  3. Hyperspatial mapping of water, energy and carbon fluxes with Unmanned Aerial Vehicles

    DEFF Research Database (Denmark)

    Wang, Sheng; Köppl, Christian Josef; Bandini, Filippo

    Having spatially distributed estimates of energy, water and carbon fluxes between the land and the atmosphere is of critical importance for improving water resource management, agricultural production, weather forecasting, and climate prediction. Traditionally, satellite based remote sensing data...... biomass, plant diseases or stress, water uptake.......Having spatially distributed estimates of energy, water and carbon fluxes between the land and the atmosphere is of critical importance for improving water resource management, agricultural production, weather forecasting, and climate prediction. Traditionally, satellite based remote sensing data...

  4. Carbon dioxide, water vapour and energy fluxes over a semi ...

    Indian Academy of Sciences (India)

    42

    Key words: Eddy covariance; semi-evergreen forest; surface energy balance; Carbon dioxide. 25 flux; Indian .... following forest types and biomes (Champion and Seth 1968), namely Eastern wet alluvial. 86 grasslands ..... Turbulence characteristics of the site were analyzed (table 2 a-b)during whole period. 197 and four ...

  5. Flux and polarisation spectra of water clouds on exoplanets

    NARCIS (Netherlands)

    Karalidi, T.; Stam, D.M.; Hovenier, J.W.

    2011-01-01

    Context. A crucial factor for a planet’s habitability is its climate. Clouds play an important role in planetary climates. Detecting and characterising clouds on an exoplanet is therefore crucial when addressing this planet’s habitability. Aims. We present calculated flux and polarisation spectra of

  6. Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept

    OpenAIRE

    Haileyesus Tsige-Tamirat; Luca Ammirabile

    2015-01-01

    Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR) has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron s...

  7. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  8. Final Report Independent Verification Survey of the High Flux Beam Reactor, Building 802 Fan House Brookhaven National Laboratory Upton, New York

    Energy Technology Data Exchange (ETDEWEB)

    Harpeneau, Evan M. [Oak Ridge Institute for Science and Education, Oak Ridge, TN (United States). Independent Environmental Assessment and Verification Program

    2011-06-24

    On May 9, 2011, ORISE conducted verification survey activities including scans, sampling, and the collection of smears of the remaining soils and off-gas pipe associated with the 802 Fan House within the HFBR (High Flux Beam Reactor) Complex at BNL. ORISE is of the opinion, based on independent scan and sample results obtained during verification activities at the HFBR 802 Fan House, that the FSS (final status survey) unit meets the applicable site cleanup objectives established for as left radiological conditions.

  9. SAFARI 2000 Kalahari Transect CO2, Water Vapor, and Heat Flux, Wet Season 2000

    Data.gov (United States)

    National Aeronautics and Space Administration — Short-term measurements of carbon dioxide, water, and energy fluxes were collected at four locations along a mean annual precipitation gradient in southern Africa...

  10. SAFARI 2000 Kalahari Transect CO2, Water Vapor, and Heat Flux, Wet Season 2000

    Data.gov (United States)

    National Aeronautics and Space Administration — ABSTRACT: Short-term measurements of carbon dioxide, water, and energy fluxes were collected at four locations along a mean annual precipitation gradient in southern...

  11. Latent heat flux measurements over complex terrain by airborne water vapour and wind lidars

    National Research Council Canada - National Science Library

    Kiemle, Christoph; Wirth, Martin; Fix, Andreas; Rahm, Stephan; Corsmeier, Ulrich; Di Girolamo, Paolo

    2011-01-01

    Vertical profiles of the latent heat flux in a convective boundary layer (CBL) are obtained for the first time over complex terrain with airborne water vapour differential absorption lidar and Doppler wind lidar...

  12. IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

    2004-10-03

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.

  13. Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach

    Energy Technology Data Exchange (ETDEWEB)

    Zaidabadinejad, Majid; Ansarifar, Gholam Reza [Isfahan Univ. (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2017-11-15

    In nuclear reactor imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, poisons concentrations, especially xenon must be accessible. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the axial xenon oscillations and ensures these oscillations are kept bounded within allowable limits during load-following operation, a robust observer based nonlinear control based on multipoint kinetics reactor model for pressurized-water nuclear reactors is presented. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermal-hydraulic). Simulation results are presented to demonstrate the effectiveness of the proposed observer based controller for the load-following operation.

  14. THE INFLUENCE OF MIEX® RESIN FOR WATER TREATMENT EFFICIENCYIN A HYBRID MEMBRANE REACTOR

    Directory of Open Access Journals (Sweden)

    Mariola Rajca

    2014-10-01

    Full Text Available The paper presents the results of studies related to the effectiveness of removal of natural organic matter (NOM from water using hybrid membrane reactor in which ion exchange and ultrafiltration processes were performed. MIEX® resin by Orica Watercare and immersed ultrafiltration polyvinylidene fluoride capillary module ZeeWeed 1 (ZW 1 by GE Power&Water operated at negative pressure were used. The application of multifunctional reactor had a positive effect on the removal of contaminants and enabled the production of high quality water. Additionally, in refer to single stage ultrafiltration it minimalized the occurrence of membrane fouling.

  15. Water distribution in a sorption enhanced methanation reactor by time resolved neutron imaging.

    Science.gov (United States)

    Borgschulte, A; Delmelle, R; Duarte, R B; Heel, A; Boillat, P; Lehmann, E

    2016-06-29

    Water adsorption enhanced catalysis has been recently shown to greatly increase the conversion yield of CO2 methanation. However, the joint catalysis and adsorption process requires new reactor concepts. We measured the spatial water distribution in a model fixed bed reactor using time resolved neutron imaging. Due to the high neutron attenuation coefficient of hydrogen, the absorbed water in the sorption catalyst gives a high contrast allowing us to follow its formation and map its distribution. At the same time, the product gas was analysed by FTIR-gas analysis. The measurements provided crucial insights into the future design of sorption reactors: during the sorption enhanced reaction, a reaction front runs through the reactor. Once the extension of the reaction front reaches the exhaust, the conversion rate of sorption enhanced methanation decreases. The existence of a reaction front running through the reactor is prerequisite for a high conversion rate. We give a simple model of the experimental results, in particular the conditions, under which a reaction front is established. In particular the latter effect must be taken into account for the dimensions of a large scale reactor.

  16. Production of medical radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for cancer treatment and arterial restenosis therapy after PTCA

    Energy Technology Data Exchange (ETDEWEB)

    Knapp, F.F. Jr.; Beets, A.L.; Mirzadeh, S.; Alexander, C.W.; Hobbs, R.L.

    1998-06-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  17. UV disinfection of indigenous aerobic spores: implications for UV reactor validation in unfiltered waters.

    Science.gov (United States)

    Mamane-Gravetz, Hadas; Linden, Karl G

    2004-07-01

    Conventional validation testing of UV reactors use cultured microorganisms spiked into test water flowing through a reactor. These tests are limited by the microbe titer it is possible to grow, thus limiting the size of the reactor it is possible to validate. The goal of this study was to examine the UV inactivation of indigenous aerobic spores naturally occurring in raw/unfiltered water supplies and to assess their use as an alternative indicator for validation testing of UV reactor performance, specifically for unfiltered water supplies planning large UV reactors. These spores were found in all raw waters tested in concentrations ranging between 20 and 12,000 CFU/100 mL and were very resistant to UV irradiation compared to a range of different microbes in the literature (i.e. adenovirus, MS-2 coliphage, and Cryptosporidium parvum). The inactivation of indigenous natural aerobic spores followed first-order kinetics with an inactivation coefficient ranging between 0.013 and 0.022 cm2/mJ with a high correlation coefficient. It was determined that naturally occurring aerobic spores, well characterized with respect to UV 253.7 nm inactivation, can be a useful tool when validating plant performance, and might also be used as a regular monitor of UV fluence and performance in a water treatment plant.

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

    Science.gov (United States)

    Corletti, Michael M.; Schulz, Terry L.

    1993-01-01

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

  19. Analysis of maximum allowable fragment heights during dissolution of high flux isotope reactor fuel in an h-canyon dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-17

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved using a flowsheet developed by the Savannah River National Laboratory (SRNL) in either the 6.4D or 6.1D dissolver using a unique insert. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The recovered U will be down-blended into low-enriched U for subsequent use as commercial reactor fuel. During the development of the HFIR fuel dissolution flowsheet, the cycle time for the initial core was estimated at 28 to 40 h. Once the cycle is complete, H-Canyon personnel will open the dissolver and probe the HFIR insert wells to determine the height of any fuel fragments which did not dissolve. Before the next core can be charged to the dissolver, an analysis of the potential for H2 gas generation must show that the combined surface area of the fuel fragments and the subsequent core will not generate H2 concentrations in the dissolver offgas which exceeds 60% of the lower flammability limit (LFL) of H2 at 200 °C. The objective of this study is to identify the maximum fuel fragment height as a function of the Al concentration in the dissolving solution which will provide criteria for charging successive HFIR cores to an H-Canyon dissolver.

  20. Removing 17β-estradiol from drinking water in a biologically active carbon (BAC) reactor modified from a granular activated carbon (GAC) reactor.

    Science.gov (United States)

    Li, Zhongtian; Dvorak, Bruce; Li, Xu

    2012-06-01

    Estrogenic compounds in drinking water sources pose potential threats to human health. Treatment technologies are needed to effectively remove these compounds for the production of safe drinking water. In this study, GAC adsorption was first tested for its ability to remove a model estrogenic compound, 17β-estradiol (E2). Although GAC showed a relatively high adsorption capacity for E2 in isotherm experiments, it appeared to have a long mass transfer zone in a GAC column reactor, causing an early leakage of E2 in the effluent. With an influent E2 concentration of 20 μg/L, the GAC reactor was able to bring down effluent E2 to ≈ 200 ng/L. To further enhance E2 removal, the GAC reactor was converted to a biologically active carbon (BAC) reactor by promoting biofilm growth in the reactor. Under optimal operating conditions, the BAC reactor had an effluent E2 concentration of ≈ 50 ng/L. With the empty bed contact times tested, the reactor exhibited more robust E2 removal performance under the BAC operation than under the GAC operation. It is noted that estrone (E1), an E2 biodegradation intermediate, was frequently detected in reactor effluent during the BAC operation. Results from this study suggested that BAC could be an effective drinking water treatment process for E2 removal and in the meantime E1 accumulation needs to be addressed. Published by Elsevier Ltd.

  1. Detection of Metabolic Fluxes of O and H Atoms into Intracellular Water in Mammalian Cells

    Science.gov (United States)

    Kreuzer, Helen W.; Quaroni, Luca; Podlesak, David W.; Zlateva, Theodora; Bollinger, Nikki; McAllister, Aaron; Lott, Michael J.; Hegg, Eric L.

    2012-01-01

    Metabolic processes result in the release and exchange of H and O atoms from organic material as well as some inorganic salts and gases. These fluxes of H and O atoms into intracellular water result in an isotopic gradient that can be measured experimentally. Using isotope ratio mass spectroscopy, we revealed that slightly over 50% of the H and O atoms in the intracellular water of exponentially-growing cultured Rat-1 fibroblasts were isotopically distinct from growth medium water. We then employed infrared spectromicroscopy to detect in real time the flux of H atoms in these same cells. Importantly, both of these techniques indicate that the H and O fluxes are dependent on metabolic processes; cells that are in lag phase or are quiescent exhibit a much smaller flux. In addition, water extracted from the muscle tissue of rats contained a population of H and O atoms that were isotopically distinct from body water, consistent with the results obtained using the cultured Rat-1 fibroblasts. Together these data demonstrate that metabolic processes produce fluxes of H and O atoms into intracellular water, and that these fluxes can be detected and measured in both cultured mammalian cells and in mammalian tissue. PMID:22848359

  2. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  3. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  4. Anaerobic Treatment of Concentrated Black Water in a UASB Reactor at a Short HRT

    Directory of Open Access Journals (Sweden)

    Cees J. N. Buisman

    2010-02-01

    Full Text Available This research describes the feasibility of applying a UASB reactor for the treatment of concentrated black (toilet water at 25 °C. On average 78% of the influent load of COD at an HRT of 8.7 days was removed. Produced methane can be converted to 56 MJ/p/y as electricity and 84 MJ/p/y as heat by combined heat and power (CHP. Minimum reactor volume at full scale was calculated to be 63L per person (for black water containing 16 gCOD/L produced at 5 L/p/d and this is more than two times smaller than other type of reactors for anaerobic treatment of concentrated black water.

  5. Benthic trophic status and nutrient fluxes in shallow-water sediments

    Science.gov (United States)

    Engelsen, Anna; Hulth, Stefan; Pihl, Leif; Sundbäck, Kristina

    2008-07-01

    Proliferation of fast-growing ephemeral macroalgae in shallow-water embayments constitutes a large-scale environmental change of coastal marine ecosystems. Since inorganic nutrients essential for the initiation and maintenance of macroalgal growth may be supplied from the underlying sediment, we investigated the coupling between benthic inorganic nutrient (mainly N and P) fluxes and sediment properties in 6 bays representing a wide gradient of sediment characteristics (grain size, organic matter content, solid phase C and N). The initial characterization of bays was made in June and also included measurements of oxygen flux and microphytobenthic and macrofaunal biomass. In September, still within the growth season of the macroalgae, complementary experiments with sediment-water incubations for benthic flux measurements of oxygen and nutrients focused on trophic status (balance between auto- and heterotrophy) as a controlling factor for rates of measured benthic nutrient fluxes. Generally, sediments rendered autotrophic by microphytobenthic photosynthesis removed nutrients from the overlying water, while heterotrophic sediments supplied nutrients to the overlying bottom water. Estimations of the green-algal nutrient demand suggested that late in the growth season, net heterotrophic sediments could cover 20% of the N-demand and 70% of the P demand. As the benthic trophic status is a functional variable more closely coupled to nutrient fluxes than the comparably conservative structural parameter organic matter content, we suggest that the trophic status is a more viable parameter to classify sediments and predict benthic nutrient fluxes in shallow-water environments.

  6. OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels.

  7. Assessing environmental impacts on stream water quality: the use of cumulative flux and cumulative flux difference approaches to deforestation of the Hafren Forest, mid-Wales

    Directory of Open Access Journals (Sweden)

    C. Neal

    2002-01-01

    Full Text Available A method for examining the impacts of disturbance on stream water quality based on paired catchment “controlâ€? and “responseâ€? water quality time series is described in relation to diagrams of cumulative flux and cumulative flux difference. The paper describes the equations used and illustrates the patterns expected for idealised flux changes followed by an application to stream water quality data for a spruce forested catchment, the Hore, subjected to clear fell. The water quality determinands examined are sodium, chloride, nitrate, calcium and acid neutralisation capacity. The anticipated effects of felling are shown in relation to reduction in mist capture and nitrate release with felling as well as to the influence of weathering and cation exchange mechanisms, but in a much clearer way than observed previously using other approaches. Keywords: Plynlimon, stream, Hore, acid neutralisation capacity, calcium, chloride, nitrate, sodium, cumulative flux, flux

  8. Applicability of a multivariable autoregressive method to boiling water reactor core stability estimation

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, S.; Fukunishi, K.; Kishi, S.; Yoshimoto, Y.; Kishimoto, K.

    1986-08-01

    A multivariable autoregressive (MAR) method is applied to the core stability estimation of a boiling water reactor-5 operation. Noise data measured during steady-state operations at startup tests are used. In this method, the closed loop transfer function from reactor pressure to reactor power is identified from reactor noise data and transformed into an impulse response function. The decay ratio representing stability characteristics is evaluated from this function. The variation range of decay ratio estimates obtained by this method is sufficiently small, if the analyzing conditions are appropriately selected. The value of the decay ratio is 0.23 during natural circulation and decreases with core flow, reaching close to zero at the rated power. A similar power dependence for the decay ratio is seen in results from a core stability analysis code. The MAR method is a useful tool for stability estimation, even if no external disturbance tests are conducted.

  9. An economic optimization of pressurized light water reactor cores

    Science.gov (United States)

    Pfeifer, Holger

    Two reactor cores (1000 MWe and 600 MWe) are optimized with respect to power cost. The power cost is minimized while retaining the thermal-hydraulic margins of the reference core. Constant thermal-hydraulic margins result in similar accident thermal-hydraulic transient behavior of the cores developed during the optimization study. The cost components impacted by the optimization are once-through fuel cycle, capital, and administrative/manpower costs. The variables in the optimization are pin diameter, moderator to fuel (H/U) ratio, core length, and the number of fuel pins in the core. A sequential quadratic programming approach is employed to solve the nonlinear optimization problem with constraints. The fuel cycle costs are evaluated by the use of the linear reactivity model, and capital costs are adjusted by suitable modifications to the nuclear energy cost database reference costs. The results of the analysis shows that for fixed assembly parameters (i.e., pin diameter, H/U ratio, and core length), the optimum core is one that operates at the thermal-hydraulic limits. Cores optimized with unconstrained assembly characteristics contain a larger number of smaller pins at a higher H/U ratio. This follows the trend in current reactor designs. While the lifetime power cost savings for the optimized core are less than 4 million dollars (versus a present day total cost of 6.9 billion dollars), the optimization analysis shows that higher thermal-hydraulic margins can be attained with minimum power cost increases. With increased emphasis on reactor safety, significantly higher safety margins may therefore be achieved without a significant power cost increase. The optimized configurations were found to be relatively insensitive to fuel cycle cost component variations.

  10. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  11. Franco-German nuclear cooperation: from the `common product` to the first European pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vignon, D. [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-la-Defense (France)

    1999-01-01

    It has now been 10 years since Framatome and Siemens decided to collaborate on the design and sales of an advanced nuclear power plant (NPP) model based on pressurized water reactor (PWR) technology. Originally called the `common product`, this model was renamed the European pressurized water reactor when Electricite de France (EDF) and the German electric utilities joined this cooperative development effort in 1992. Since the beginning, this cooperation has been formalized in the framework of an agreement that led to the founding of a joint and equally owned subsidiary, Nucler Power International (NPI), which is reponsible for leading the development of the new model and later handling its export sales.

  12. Estimation of In Situ Sediment-to-Water Fluxes of Polycyclic Aromatic Hydrocarbons, Polychlorobiphenyls and Polybrominated Diphenylethers

    NARCIS (Netherlands)

    Koelmans, A.A.; Poot, A.; Lange, de H.J.; Velzeboer, I.; Harmsen, J.; Noort, van P.C.M.

    2010-01-01

    Sediment--water fluxes of hydrophobic organic chemicals (HOC) may affect the quality of surface waters. Here, we present an approach to derive such fluxes from (a) in situ HOC concentration gradients measured with passive samplers and (b) mass transfer coefficients measured with a novel flux method

  13. Production of medical radioisotopes in the ORNL high flux isotope reactor (HFIR) for cancer treatment and arterial restenosis therapy after PICA

    Science.gov (United States)

    Knapp, F. F.; Beets, A. L.; Mirzadeh, S.; Alexander, C. W.; Hobbs, R. L.

    1999-01-01

    The High Flux Isotope Reactor ( HFIR) at the Oak Ridge National Laboratory ( ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. First beginning operation in 1965, the high thermal neutron flux (2.5×1015 neutrons/cm2/sec at 85 MW) and versatile target irradiation and handling facilities provide the opportunity for production of a wide variety of neutron-rich medical radioisotopes of current interest for therapy. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117 m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube ( HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle (22-24 days) and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions ( PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117 m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  14. Modeling and Predicting Carbon and Water Fluxes Using Data-Driven Techniques in a Forest Ecosystem

    Directory of Open Access Journals (Sweden)

    Xianming Dou

    2017-12-01

    Full Text Available Accurate estimation of carbon and water fluxes of forest ecosystems is of particular importance for addressing the problems originating from global environmental change, and providing helpful information about carbon and water content for analyzing and diagnosing past and future climate change. The main focus of the current work was to investigate the feasibility of four comparatively new methods, including generalized regression neural network, group method of data handling (GMDH, extreme learning machine and adaptive neuro-fuzzy inference system (ANFIS, for elucidating the carbon and water fluxes in a forest ecosystem. A comparison was made between these models and two widely used data-driven models, artificial neural network (ANN and support vector machine (SVM. All the models were evaluated based on the following statistical indices: coefficient of determination, Nash-Sutcliffe efficiency, root mean square error and mean absolute error. Results indicated that the data-driven models are capable of accounting for most variance in each flux with the limited meteorological variables. The ANN model provided the best estimates for gross primary productivity (GPP and net ecosystem exchange (NEE, while the ANFIS model achieved the best for ecosystem respiration (R, indicating that no single model was consistently superior to others for the carbon flux prediction. In addition, the GMDH model consistently produced somewhat worse results for all the carbon flux and evapotranspiration (ET estimations. On the whole, among the carbon and water fluxes, all the models produced similar highly satisfactory accuracy for GPP, R and ET fluxes, and did a reasonable job of reproducing the eddy covariance NEE. Based on these findings, it was concluded that these advanced models are promising alternatives to ANN and SVM for estimating the terrestrial carbon and water fluxes.

  15. Local stability tests in Dresden 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

  16. On the study of catalytic membrane reactor for water detritiation: Modeling approach

    Energy Technology Data Exchange (ETDEWEB)

    Liger, Karine, E-mail: karine.liger@cea.fr [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Mascarade, Jérémy [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Joulia, Xavier; Meyer, Xuan-Mi [Université de Toulouse, INPT, UPS, Laboratoire de Génie Chimique, 4, Allée Emile Monso, Toulouse F-31030 (France); CNRS, Laboratoire de Génie Chimique, Toulouse F-31030 (France); Troulay, Michèle; Perrais, Christophe [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France)

    2016-11-01

    Highlights: • Experimental results for the conversion of tritiated water (using deuterium as a simulant of tritium) by means of a catalytic membrane reactor in view of tritium recovery. • Phenomenological 2D model to represent catalytic membrane reactor behavior including the determination of the compositions of gaseous effluents. • Good agreement between the simulation results and experimental measurements performed on the dedicated facility. • Explanation of the unexpected behavior of the catalytic membrane reactor by the modeling results and in particular the gas composition estimation. - Abstract: In the framework of tritium recovery from tritiated water, efficiency of packed bed membrane reactors have been successfully demonstrated. Thanks to protium isotope swamping, tritium bonded water can be recovered under the valuable Q{sub 2} form (Q = H, D or T) by means of isotope exchange reactions occurring on catalyst surface. The use of permselective Pd-based membrane allows withdrawal of reactions products all along the reactor, and thus limits reverse reaction rate to the benefit of the direct one (shift effect). The reactions kinetics, which are still little known or unknown, are generally assumed to be largely greater than the permeation ones so that thermodynamic equilibriums of isotope exchange reactions are generally assumed. This paper proposes a new phenomenological 2D model to represent catalytic membrane reactor behavior with the determination of gas effluents compositions. A good agreement was obtained between the simulation results and experimental measurements performed on a dedicated facility. Furthermore, the gas composition estimation permits to interpret unexpected behavior of the catalytic membrane reactor. In the next future, further sensitivity analysis will be performed to determine the limits of the model and a kinetics study will be conducted to assess the thermodynamic equilibrium of reactions.

  17. Validation of effective momentum and heat flux models for stratification and mixing in a water pool

    Energy Technology Data Exchange (ETDEWEB)

    Hua Li; Villanueva, W.; Kudinov, P. [Royal Institute of Technology (KTH), Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-06-15

    The pressure suppression pool is the most important feature of the pressure suppression system in a Boiling Water Reactor (BWR) that acts primarily as a passive heat sink during a loss of coolant accident (LOCA) or when the reactor is isolated from the main heat sink. The steam injection into the pool through the blowdown pipes can lead to short term dynamic phenomena and long term thermal transient in the pool. The development of thermal stratification or mixing in the pool is a transient phenomenon that can influence the pool's pressure suppression capacity. Different condensation regimes depending on the pool's bulk temperature and steam flow rates determine the onset of thermal stratification or erosion of stratified layers. Previously, we have proposed to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) and the Effective Momentum Source (EMS) models. The EHS model is used to provide thermal effect of steam injection on the pool, preserving heat and mass balance. The EMS model is used to simulate momentum induced by steam injection in different flow regimes. The EMS model is based on the combination of (i) synthetic jet theory, which predicts effective momentum if amplitude and frequency of flow oscillations in the pipe are given, and (ii) model proposed by Aya and Nariai for prediction of the amplitude and frequency of oscillations at a given pool temperature and steam mass flux. The complete EHS/EMS models only require the steam mass flux, initial pool bulk temperature, and design-specific parameters, to predict thermal stratification and mixing in a pressure suppression pool. In this work we use EHS/EMS models implemented in containment thermal hydraulic code GOTHIC. The PPOOLEX experiments (Lappeenranta University of Technology, Finland) are utilized to (a) quantify errors due to GOTHIC's physical models and numerical schemes, (b) propose necessary improvements in GOTHIC sub-grid scale

  18. Surface fluxes and water balance of spatially varying vegetation within a small mountainous headwater catchment

    Directory of Open Access Journals (Sweden)

    G. N. Flerchinger

    2010-06-01

    Full Text Available Precipitation variability and complex topography often create a mosaic of vegetation communities in mountainous headwater catchments, creating a challenge for measuring and interpreting energy and mass fluxes. Understanding the role of these communities in modulating energy, water and carbon fluxes is critical to quantifying the variability in energy, carbon, and water balances across landscapes. The focus of this paper was: (1 to demonstrate the utility of eddy covariance (EC systems in estimating the evapotranspiration component of the water balance of complex headwater mountain catchments; and (2 to compare and contrast the seasonal surface energy and carbon fluxes across a headwater catchment characterized by large variability in precipitation and vegetation cover. Eddy covariance systems were used to measure surface fluxes over sagebrush (Artemesia arbuscula and Artemesia tridentada vaseyana, aspen (Populus tremuloides and the understory of grasses and forbs beneath the aspen canopy. Peak leaf area index of the sagebrush, aspen, and aspen understory was 0.77, 1.35, and 1.20, respectively. The sagebrush and aspen canopies were subject to similar meteorological forces, while the understory of the aspen was sheltered from the wind. Missing periods of measured data were common and made it necessary to extrapolate measured fluxes to the missing periods using a combination of measured and simulated data. Estimated cumulative evapotranspiratation from the sagebrush, aspen trees, and aspen understory were 384 mm, 314 mm and 185 mm. A water balance of the catchment indicated that of the 699 mm of areal average precipitation, 421 mm was lost to evapotranspiration, and 254 mm of streamflow was measured from the catchment; water balance closure for the catchment was within 22 mm. Fluxes of latent heat and carbon for all sites were minimal through the winter. Growing season fluxes of latent heat and carbon were consistently higher

  19. Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Zuoyi [Inst. of Nuclear Energy Technology Tsinghua Univ., Beijing, BJ (China); Scherer, Winfried

    1997-12-31

    This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H{sub 2}O density increase rate in the reactor core is smaller than 0.3 kg/(m{sup 3}s). The liquid water vaporization in the steam generator and H{sub 2}O transport from the steam generator to the reactor core reduces the impulse of the H{sub 2}O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

  20. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [and others

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  1. Improving SWAT for simulating water and carbon fluxes of forest ecosystems

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Qichun; Zhang, Xuesong

    2016-11-01

    As a widely used watershed model for assessing impacts of anthropogenic and natural disturbances on water quantity and quality, the Soil and Water Assessment Tool (SWAT) has not been extensively tested in simulating water and carbon fluxes of forest ecosystems. Here, we examine SWAT simulations of evapotranspiration (ET), net primary productivity (NPP), net ecosystem exchange (NEE), and plant biomass at ten AmeriFlux forest sites across the U.S. We identify unrealistic radiation use efficiency (Bio_E), large leaf to biomass fraction (Bio_LEAF), and missing phosphorus supply from parent material weathering as the primary causes for the inadequate performance of the default SWAT model in simulating forest dynamics. By further revising the relevant parameters and processes, SWAT’s performance is substantially improved. Based on the comparison between the improved SWAT simulations and flux tower observations, we discuss future research directions for further enhancing model parameterization and representation of water and carbon cycling for forests.

  2. DEVELOPMENT AND DEMONSTRATION OF A BIDIRECTIONAL ADVECTIVE FLUX METER FOR SEDIMENT-WATER INTERFACE

    Science.gov (United States)

    A bidirectional advective flux meter for measuring water transport across the sediment-water interface has been successfully developed and field tested. The flow sensor employs a heat-pulse technique combined with a flow collection funnel for the flow measurement. Because the dir...

  3. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gavilian-Moreno, Carlos [Iberdrola Generacion, S.A., Cofrentes Nuclear Power Plant, Project Engineering Department, Paraje le Plano S/N, Valencia (Spain); Espinosa-Paredes, Gilberto [Area de ingeniera en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Mexico city (Mexico)

    2016-04-15

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  4. Using Largest Lyapunov Exponent to Confirm the Intrinsic Stability of Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Carlos J. Gavilán-Moreno

    2016-04-01

    Full Text Available The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs. Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  5. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  6. Stratigraphic controls on fluid and solute fluxes across the sediment-water interface of an estuary

    Science.gov (United States)

    Sawyer, Audrey H.; Lazareva, Olesya; Kroeger, Kevin D.; Crespo, Kyle; Chan, Clara S.; Stieglitz, Thomas; Michael, Holly A.

    2014-01-01

    Shallow stratigraphic features, such as infilled paleovalleys, modify fresh groundwater discharge to coastal waters and fluxes of saltwater and nutrients across the sediment–water interface. We quantify the spatial distribution of shallow surface water–groundwater exchange and nitrogen fluxes near a paleovalley in Indian River Bay, Delaware, using a hand resistivity probe, conventional seepage meters, and pore-water samples. In the interfluve (region outside the paleovalley) most nitrate-rich fresh groundwater discharges rapidly near the coast with little mixing of saline pore water, and nitrogen transport is largely conservative. In the peat-filled paleovalley, fresh groundwater discharge is negligible, and saltwater exchange is deep (∼1 m). Long pore-water residence times and abundant sulfate and organic matter promote sulfate reduction and ammonium production in shallow sediment. Reducing, iron-rich fresh groundwater beneath paleovalley peat discharges diffusely around paleovalley margins offshore. In this zone of diffuse fresh groundwater discharge, saltwater exchange and dispersion are enhanced, ammonium is produced in shallow sediments, and fluxes of ammonium to surface water are large. By modifying patterns of groundwater discharge and the nature of saltwater exchange in shallow sediments, paleovalleys and other stratigraphic features influence the geochemistry of discharging groundwater. Redox reactions near the sediment–water interface affect rates and patterns of geochemical fluxes to coastal surface waters. For example, at this site, more than 99% of the groundwater-borne nitrate flux to the Delaware Inland Bays occurs within the interfluve portion of the coastline, and more than 50% of the ammonium flux occurs at the paleovalley margin.

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

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

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

  8. Modelling of gas flux through bubbles at the air-water interface

    OpenAIRE

    Memery, Laurent; Merlivat, Liliane

    2011-01-01

    A theoretical approach to gas transfer by bubbles created by breaking waves at the air-waterinterface is undertaken. Based on a simple model, this study affords a basic understanding of thephysical mechanisms. The behaviour of a single bubble is examined. Similarities and differencesbetween the formulation of gas fluxes across a “flat” air-water interface and through bubbles areshown. The gas flux through bubbles is not strictly proportional to the solubility and to thedifference of concentra...

  9. Passive methods for quantifying the In Situ Flux of Water, Uranium, and Microbial Biomass

    Science.gov (United States)

    Newman, M. A.; Peacock, A.; Hatfield, K.; Stucker, V.; Cho, J.; Klammler, H.; Ranville, J. F.; Cabaniss, S.; Annable, M. D.; Perminova, I.

    2011-12-01

    The goal of this project was to develop a novel sensor that incorporates field-tested concepts of the passive flux meter (PFM) to provide direct in situ measures of uranium and groundwater fluxes. The sensor uses two sorbents and tracers to measure uranium flux and specific discharge directly-sensor principles and design will apply to fluxes of other radionuclides, metals, and co-contaminants. Flux measurements will assist in obtaining field-scale quantification of subsurface processes affecting uranium transport (e.g., advection) and transformation (e.g., uranium attenuation) and further advance conceptual and computational models for field scale simulations. Project efforts will expand our current understanding of how field-scale spatial variations in fluxes of uranium, groundwater and salient electron donor/acceptors are coupled to spatial variations in measured microbial biomass/community composition, effective field-scale uranium mass balances, attenuation, and stability. Field tests in the La Quinta and Super 8 galleries at the Rifle IFRC site were conducted to assess ambient groundwater, uranium, and microbial biomass fluxes. The latter were determined using a newly designed Baffled Multilevel Sampling (BMLS) device installed in typical screened monitoring wells to provide aqueous concentrations of dissolved or suspended constituents over multiple isolated vertical sections of the well. Biomass mass fluxes were calculated from the product of BMLS data for microbial cell counts from PCR analyses and PFM water fluxes collected from coincident well sections. Expected microbial discharge for Eubacteria in the La Quinta gallery was estimated to be 1.7 x 1012 cells per day. The biomass discharges for Geobacter, Methanogens, and Anaeromyxobacter remain to be determined. Expected uranium discharges predicted from stochastic simulations using PFM measures of flux over the La Quinta gallery transect and the injection-well transect of the Super 8 gallery were 26 mg

  10. Particle Filter-Based Recursive Data Fusion With Sensor Indexing for Large Core Neutron Flux Estimation

    Science.gov (United States)

    Tamboli, Prakash Kumar; Duttagupta, Siddhartha P.; Roy, Kallol

    2017-06-01

    We introduce a sequential importance sampling particle filter (PF)-based multisensor multivariate nonlinear estimator for estimating the in-core neutron flux distribution for pressurized heavy water reactor core. Many critical applications such as reactor protection and control rely upon neutron flux information, and thus their reliability is of utmost importance. The point kinetic model based on neutron transport conveniently explains the dynamics of nuclear reactor. The neutron flux in the large core loosely coupled reactor is sensed by multiple sensors measuring point fluxes located at various locations inside the reactor core. The flux values are coupled to each other through diffusion equation. The coupling facilitates redundancy in the information. It is shown that multiple independent data about the localized flux can be fused together to enhance the estimation accuracy to a great extent. We also propose the sensor anomaly handling feature in multisensor PF to maintain the estimation process even when the sensor is faulty or generates data anomaly.

  11. Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mkrtchyan, Lilit; Schau, Henry [TUEV SUED Energietechnik GmbH, Mannheim (Germany). Abt. Strukturverhalten; Wolf, Werner; Holzer, Wieland [TUEV SUED Industrie Service GmbH, Muenchen (Germany). Abt. Behaelter und Turbosatz; Wernicke, Robert; Trieglaff, Ralf [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany). Abt. Festigkeit und Konstruktion

    2011-08-15

    The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

  12. A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.Y.; Bentz, J.; Simpson, R. [Sandia National Labs., Albuquerque, NM (United States); Witt, R. [Univ. of Wisconsin, Madison, WI (United States)

    1997-02-01

    The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF program investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.

  13. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

  14. TMI-2 Reactor Building source term measurements: surfaces and basement water and sediment

    Energy Technology Data Exchange (ETDEWEB)

    McIsaac, C V; Keefer, D G

    1984-10-01

    Presented in this report are the results of radiochemical and elemental analyses performed on samples collected from the Three Mile Island Unit 2 Reactor Building from August 1979 to December 1983. The quantities of fission products and core materials that were measured on the external surfaces in the Reactor Building or in the water and sediment in its basement are summarized. Recent analysis results for access panels removed from the air cooling assembly and for liquid and particulate samples collected from the Reactor Building sump and reactor coolant drain tank are included in the report. Measurements show that 59% of the /sup 3/H, 2.7% of the /sup 90/Sr, 15% of the /sup 129/I, 20% of the /sup 131/I, and 42% of the /sup 137/Cs originally in the core at the time of the accident could be accounted for outside the core in the Reactor Building. With the exceptions of /sup 90/Sr and /sup 144/Ce, the vast majority of each radionuclide released was found dispersed in the water and sediment in the basement.

  15. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

  16. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition,

  17. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-03-20

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... license renewal interim staff guidance (LR-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures and...

  18. Light-water-reactor safety program. Quarterly progress report, April--June 1977

    Energy Technology Data Exchange (ETDEWEB)

    Sachs, R G; Kyger, J A

    1977-01-01

    The report summarizes work performed on the following water-reactor-safety problems: (1) loss-of-coolant accident research in heat transfer and fluid dynamics; (2) transient fuel response and fission-product release; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

  19. Local Fission Gas Release and Swelling in Water Reactor Fuel during Slow Power Transients

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg; Walker, C.T.; Ray, I.L.F.

    1985-01-01

    Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7–4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25–40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two...

  20. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  1. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

  2. Study of water radiolysis in relation with the primary cooling circuit of pressurized water reactors; Etude sur la radiolyse de l`eau en relation avec le circuit primaire de refroidissement des reacteurs nucleaires a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Pastina, B

    1997-07-01

    This memorandum shows a fundamental study on the water radiolysis in relation with the cooling primary circuit of PWR type reactors. The water of the primary circuit contains boric acid a soluble neutronic poison and also hydrogen that has for role to inhibit the water decomposition under radiation effect. In the aim to better understand the mechanism of dissolved hydrogen action and to evaluate the impact of several parameters on this mechanism, aqueous solutions with boric acid and hydrogen have been irradiated in a experimental nuclear reactor, at 30, 100 and 200 Celsius degrees. It has been found that, with hydrogen, the water decomposition under irradiation is a threshold phenomenon in function of the ratio between the radiation flux `1` B(n, )`7 Li and the gamma flux. When this ratio become too high, the number of radicals is not sufficient to participate at the chain reaction, and then water is decomposed in O{sub 2} and H{sub 2}O{sub 2} in a irreversible way. The temperature has a beneficial part on this mechanism. The iron ion and the copper ion favour the water decomposition. (N.C.). 83 refs.

  3. Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstance of NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Ho; Kim, Dae Seop; Kim, Jae Hwan; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2014-06-15

    Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

  4. Stable water isotope and surface heat flux simulation using ISOLSM: Evaluation against in-situ measurements

    KAUST Repository

    Cai, Mick Y.

    2015-04-01

    The stable isotopes of water are useful tracers of water sources and hydrological processes. Stable water isotope-enabled land surface modeling is a relatively new approach for characterizing the hydrological cycle, providing spatial and temporal variability for a number of hydrological processes. At the land surface, the integration of stable water isotopes with other meteorological measurements can assist in constraining surface heat flux estimates and discriminate between evaporation (E) and transpiration (T). However, research in this area has traditionally been limited by a lack of continuous in-situ isotopic observations. Here, the National Centre for Atmospheric Research stable isotope-enabled Land Surface Model (ISOLSM) is used to simulate the water and energy fluxes and stable water isotope variations. The model was run for a period of one month with meteorological data collected from a coastal sub-tropical site near Sydney, Australia. The modeled energy fluxes (latent heat and sensible heat) agreed reasonably well with eddy covariance observations, indicating that ISOLSM has the capacity to reproduce observed flux behavior. Comparison of modeled isotopic compositions of evapotranspiration (ET) against in-situ Fourier Transform Infrared spectroscopy (FTIR) measured bulk water vapor isotopic data (10. m above the ground), however, showed differences in magnitude and temporal patterns. The disparity is due to a small contribution from local ET fluxes to atmospheric boundary layer water vapor (~1% based on calculations using ideal gas law) relative to that advected from the ocean for this particular site. Using ISOLSM simulation, the ET was partitioned into E and T with 70% being T. We also identified that soil water from different soil layers affected T and E differently based on the simulated soil isotopic patterns, which reflects the internal working of ISOLSM. These results highlighted the capacity of using the isotope-enabled models to discriminate

  5. Source term attenuation by water in the Mark I boiling water reactor drywell

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A. [Sandia National Labs., Albuquerque, NM (United States)

    1993-09-01

    Mechanistic models of aerosol decontamination by an overlying water pool during core debris/concrete interactions and spray removal of aerosols from a Mark I drywell atmosphere are developed. Eighteen uncertain features of the pool decontamination model and 19 uncertain features of the model for the rate coefficient of spray removal of aerosols are identified. Ranges for values of parameters that characterize these uncertain features of the models are established. Probability density functions for values within these ranges are assigned according to a set of rules. A Monte Carlo uncertainty analysis of the decontamination factor produced by water pools 30 and 50 cm deep and subcooled 0--70 K is performed. An uncertainty analysis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25, 0.01, and 0.001 cm{sup 3} H{sub 2}O/cm{sup 2}-s and decontamination factors of 1.1, 2, 3.3, 10, 100, and 1000.

  6. Ammonium recovery from reject water combined with hydrogen production in a bioelectrochemical reactor.

    Science.gov (United States)

    Wu, Xue; Modin, Oskar

    2013-10-01

    In this study, a bioelectrochemical reactor was investigated for simultaneous hydrogen production and ammonium recovery from reject water, which is an ammonium-rich side-stream produced from sludge treatment processes at wastewater treatment plants. In the anode chamber of the reactor, microorganisms converted organic material into electrical current. The electrical current was used to generate hydrogen gas at the cathode with 96±6% efficiency. Real or synthetic reject water was fed to the cathode chamber where proton reduction into hydrogen gas resulted in a pH increase which led to ammonium being converted into volatile ammonia. The ammonia could be stripped from the solution and recovered in acid. Overall, ammonium recovery efficiencies reached 94% with synthetic reject water and 79% with real reject water. This process could potentially be used to make wastewater treatment plants more resource-efficient and further research is warranted. Copyright © 2013 Elsevier Ltd. All rights reserved.

  7. Carbon dioxide fluxes associated with synoptic weather events over a southern inland water

    Science.gov (United States)

    Liu, H.; Zhang, Q.; Gao, Z.

    2015-12-01

    Evidence indicates that inland waters play an important role in regional and global carbon budget through releasing a substantial carbon into the atmosphere. To better quantify how environmental variables affect CO2 exchange between inland waters and the atmosphere and its temporal variations, we have conducted direct, long-term measurements of CO2 fluxes across the water-atmosphere interface over a large southern open water of Ross Barnett Reservoir in central Mississippi. Our data indicate that large CO2 flux pulses occurred occasionally throughout the course of a year with the duration of a few days for each pulse. Here we analyzed and demonstrated that these CO2 flux pulses were associated with the passages of synoptic weather events. Our preliminary results indicated that these synoptic weather events (e.g., extratropical clones and cold air bursts) led to the enhanced mechanical mixing due to increasing wind speeds and the instability of the atmospheric surface layer due to the decreasing air temperature. As a consequence, in-water processes were also substantially altered accordingly. Due to the dramatic decrease in air temperature caused by the events, the temperature in the water surface layer was largely reduced, generating in-water convection conditions and thus leading to the increased depths of the mixing layer in the water, as reflected by the water temperature profiles. The enhanced mechanical mixing in the atmospheric surface layer may have further contributed to the deepened mixing layer in the water. Our suggestions suggest that high CO2 effluxes during the pulse events were largely attributed to changes in the water-side physical processes that are directly linked to rapid changes in atmospheric processes associated with synoptic weather events. Given its substantial contribution of CO2 flux pulses to carbon emission, such physical processes should be taken into account when carbon emissions from inland waters are quantified.

  8. Energy and Water Fluxes across a Heterogeneous Landscape in the Southern Great Plains

    Science.gov (United States)

    Bagley, J. E.; Williams, I. N.; Kueppers, L. M.; Lu, Y.; Torn, M. S.; Biraud, S.

    2015-12-01

    Fluxes of energy and water between the atmosphere and the land surface influence weather and climate. These fluxes depend on the state of the landscape, which contributes to differences in land-atmosphere coupling strength over space and time. One region with potentially strong land-atmosphere coupling is the Southern Great Plains (SGP) in North America. In this region, managed vegetation plays a key role in moderating the surface energy through effects on surface albedo, transpiration, precipitation interception, and other surface properties. However accurately modeling these effects is challenging because the vegetation in this region is very heterogeneous. Winter wheat is the dominant crop, but pasture, hayfields, corn, and recently introduced crops such as canola cover significant portions of the landscape as well. Winter wheat has a unique phenology with fall planting, maximum leaf area in late spring, and harvest in early summer. This phenology contrasts significantly with most other crops and with pastures and hayfields in the region, which have more typical spring-fall growing seasons. Therefore, to sufficiently model and assess land-atmosphere interactions in this region accurate characterization of differences in the seasonality of water and energy fluxes between vegetation types are necessary. We used observations including eddy covariance flux estimates, soil moisture data, state-of-the-art longwave and shortwave radiation measurements, and other observations available for several facilities within the SGP Atmospheric Radiation Measurement (ARM) site in north-central Oklahoma and southern Kansas. We compared the timing and variations in fluxes of water and energy between winter wheat and other land cover types, focusing on vegetation influences on rates of soil dry-down following precipitation events. We found distinct differences in fluxes between winter wheat and other land types. These flux differences had a nonlinear dependency on disparities in

  9. Design of a supercritical water-cooled reactor with a three-pass core arrangement

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, K. [EnBW Kernkraft GmbH, Kernkraftwerk Philippsburg, D-76661 Philippsburg (Germany)], E-mail: kai-fischer@gmx.de; Schulenberg, T. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, P.O. Box 3640, D-76021 Karlsruhe (Germany); Laurien, E. [University of Stuttgart, Institute for Nuclear and Energy Systems (IKE), Pfaffenwaldring 31, D-70569 Stuttgart (Germany)

    2009-04-15

    The Supercritical Water-cooled Reactor (SCWR) is one of the six concepts of the Generation IV International Forum. In Europe, investigations have been integrated into a joint research project, called High Performance Light Water Reactor (HPLWR). Due to the higher heat up within the core and a higher outlet temperature, a significant increase in turbine power and thermal efficiency of the plant can be expected. Besides the higher pressure and higher steam temperature, the design concept of this type of reactor differs significantly from a conventional LWR by a different core concept. In order to achieve the high outlet temperature of over 500 deg. C, a core with a three-step heat up and intermediate mixing is proposed to keep local cladding temperatures within today's material limits. A design for the reactor pressure vessel (RPV) and the internals has been worked out to incorporate a core arrangement with three passes. All components have been dimensioned following the safety standards of the nuclear safety standards commission in Germany. Additionally, a fuel assembly cluster with head and foot piece has been developed to facilitate the complex flow path for the multi-pass concept. The design of the internals and of the RPV is verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Furthermore, the reactor design ensures that the total coolant flow path remains closed against leakage of colder moderator water even in case of large thermal expansions of the components. The design of the RPV and internals is now available for detailed analyses of the core and the reactor.

  10. Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor

    Energy Technology Data Exchange (ETDEWEB)

    Polo-Lopez, M.I., E-mail: mpolo@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Fernandez-Ibanez, P., E-mail: pilar.fernandez@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Ubomba-Jaswa, E., E-mail: euniceubombajaswa@yahoo.com [Natural Resources and the Environment, CSIR, PO Box 395, Pretoria (South Africa); Navntoft, C., E-mail: christian.navntoft@solarmate.com.ar [Instituto de Investigacion e Ingenieria Ambiental, Universidad Nacional de San Martin (3iA-UNSAM), Peatonal Belgrano 3563, B1650ANQ San Martin (Argentina); Universidad Tecnologica Nacional - Facultad Regional Buenos Aires - Departamento de Ingenieria Civil - Laboratorio de Estudios sobre Energia Solar, (UTN-FRBA-LESES), Mozart 2300, (1407) Ciudad Autonoma de Buenos Aires, Republica Argentina (Argentina); Garcia-Fernandez, I., E-mail: irene.garcia@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Dunlop, P.S.M., E-mail: psm.dunlop@ulster.ac.uk [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); Schmid, M. [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); Byrne, J.A., E-mail: j.byrne@ulster.ac.uk [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); and others

    2011-11-30

    Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input.

  11. Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor.

    Science.gov (United States)

    Polo-López, M I; Fernández-Ibáñez, P; Ubomba-Jaswa, E; Navntoft, C; García-Fernández, I; Dunlop, P S M; Schmid, M; Byrne, J A; McGuigan, K G

    2011-11-30

    Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input. Copyright © 2011 Elsevier B.V. All rights reserved.

  12. Impacts of soil–aquifer heat and water fluxes on simulated global climate

    Directory of Open Access Journals (Sweden)

    N. Y. Krakauer

    2013-05-01

    Full Text Available Climate models have traditionally only represented heat and water fluxes within relatively shallow soil layers, but there is increasing interest in the possible role of heat and water exchanges with the deeper subsurface. Here, we integrate an idealized 50 m deep aquifer into the land surface module of the GISS ModelE general circulation model to test the influence of aquifer–soil moisture and heat exchanges on climate variables. We evaluate the impact on the modeled climate of aquifer–soil heat and water fluxes separately, as well as in combination. The addition of the aquifer to ModelE has limited impact on annual-mean climate, with little change in global mean land temperature, precipitation, or evaporation. The seasonal amplitude of deep soil temperature is strongly damped by the soil–aquifer heat flux. This not only improves the model representation of permafrost area but propagates to the surface, resulting in an increase in the seasonal amplitude of surface air temperature of > 1 K in the Arctic. The soil–aquifer water and heat fluxes both slightly decrease interannual variability in soil moisture and in land-surface temperature, and decrease the soil moisture memory of the land surface on seasonal to annual timescales. The results of this experiment suggest that deepening the modeled land surface, compared to modeling only a shallower soil column with a no-flux bottom boundary condition, has limited impact on mean climate but does affect seasonality and interannual persistence.

  13. Non normal modal analysis of oscillations in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suarez-Antola, Roberto, E-mail: roberto.suarez@miem.gub.uy [Ministerio de Industria, Energia y Mineria (MIEM), Montevideo (Uruguay); Flores-Godoy, Jose-Job, E-mail: job.flores@ibero.mx [Universidad Iberoamericana (UIA), Mexico, DF (Mexico). Dept. de Fisica Y Matematicas

    2013-07-01

    The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and

  14. Draft layout, containment and performance of the safety system of the European Supercritical Water-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schlagenhaufer, M.; Kohly, C.; Schulenberg, T. [Karlsruhe Inst. of Tech., Karlsruhe (Germany); Rothschmitt, S.; Bittermann, D. [AREVA NP GmbH, Erlangen (Germany)

    2010-07-01

    In Europe, the research on Supercritical Water-Cooled Reactors is integrated in a project called 'High Performance Light Water Reactor Phase 2' (HPLWR Phase 2), co-funded by the European Commission. Ten partners and three active supporters are working on critical scientific issues to determine the potential of this reactor concept in the electricity market. Close to the end of the project the technical results are translated into a draft layout of the HPLWR. The containment and safety system are being explained. Exemplarily, a depressurization event shows the capabilities of the safety system to sufficiently cool the reactor by means of a low pressure coolant injection system. (author)

  15. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  16. Overview of Prevention for Water Hammer by Check Valve Action in Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dayong; Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonhoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Water hammer can cause serious damage to pumping system and unexpected system pressure rise in the pipeline. In nuclear reactor, water hammer can influence on the integrity of safety related system. Water hammer in nuclear reactor have been caused by voiding in normally water-filled lines, steam condensation line containing both steam and water, as well as by rapid check valve action. Therefore, this study focuses on the water hammer by check valve among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. This study focuses on the water hammer by check valve action among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. If the inadvertent pump trip or pipe rupture in high velocity and pressure pipe is predicted, the fast response check valve such as tiled disc, dual disc and nozzle check valve should be installed in the system. If the inadvertent pump trip or pipe rupture in very high velocity and pressure pipe and excessively large revered flow velocity are predicted, the very slowly closing check valve such as controlled closure check valve should be installed in the system.

  17. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    OpenAIRE

    Ma, Weimin; Yuan, Yidan; Sehgal, Bal Raj

    2016-01-01

    A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential...

  18. Nitrates in Groundwater Discharges from the Azores Archipelago: Occurrence and Fluxes to Coastal Waters

    Directory of Open Access Journals (Sweden)

    J. Virgílio Cruz

    2017-02-01

    Full Text Available Groundwater discharge is an important vector of chemical fluxes to the ocean environment, and as the concentration of nutrients is often higher in discharging groundwater, the deterioration of water quality in the receiving environment can be the result. The main objective of the present paper is to estimate the total NO3 flux to coastal water bodies due to groundwater discharge in the volcanic Azores archipelago (Portugal. Therefore, 78 springs discharging from perched-water bodies have been monitored since 2003, corresponding to cold (mean = 14.9 °C and low mineralized (47.2–583 µS/cm groundwater from the sodium-bicarbonate to sodium-chloride water types. A set of 36 wells was also monitored, presenting groundwater with a higher mineralization. The nitrate content in springs range between 0.02 and 37.4 mg/L, and the most enriched samples are associated to the impact of agricultural activities. The total groundwater NO3 flux to the ocean is estimated in the range of 5.23 × 103 to 190.6 × 103 mol/km2/a (∑ = ~523 × 103 mol/km2/a, exceeding the total flux associated to surface runoff (∑ = ~281 × 103 mol/km2/a. In the majority of the islands, the estimated fluxes are higher than runoff fluxes, with the exception of Pico (47.2%, Corvo (46% and Faial (7.2%. The total N-NO3 flux estimated in the Azores (~118.9 × 103 mol/km2/a is in the lower range of estimates made in other volcanic islands.

  19. Measurement of the thermal and fast neutron flux in a research reactor with a Li and Th loaded optical fibre detector

    CERN Document Server

    Yamane, Y; Misawa, T; Karlsson, J K H; Pázsit, I

    1999-01-01

    The spatial dependence of thermal and fast neutron flux was measured axially in the core of a 1 MW research reactor. The measurements were made by a thin optical fibre detector with a neutron sensitive ZnS(Ag) scintillation tip. For thermal neutrons sup 6 Li was used, whereas for fast neutrons sup 2 sup 3 sup 2 Th was used as neutron converter. The spatial dependence was measured by moving the fibre axially with a uniform speed. The measurement takes a few minutes, compared to up to 10 h with the conventional wire activation method. Comparison with traditional measurements shows a good agreement. (author)

  20. Measurement of the thermal and fast neutron flux in a research reactor with a Li and Th loaded optical fibre detector

    Science.gov (United States)

    Yamane, Y.; Uritani, A.; Misawa, T.; Karlsson, J. K.-H. J. K.-H.; Pázsit, I.

    1999-08-01

    The spatial dependence of thermal and fast neutron flux was measured axially in the core of a 1 MW research reactor. The measurements were made by a thin optical fibre detector with a neutron sensitive ZnS(Ag) scintillation tip. For thermal neutrons 6Li was used, whereas for fast neutrons 232Th was used as neutron converter. The spatial dependence was measured by moving the fibre axially with a uniform speed. The measurement takes a few minutes, compared to up to 10 h with the conventional wire activation method. Comparison with traditional measurements shows a good agreement.

  1. An organic profile of a pressurised water reactor secondary plant

    Energy Technology Data Exchange (ETDEWEB)

    Eeden, Nestor van; Stwayi, Mandisibuntu; Gericke, Gerhard [Eskom Holdings SOC Ltd., Western Cape (South Africa). Koeberg Power Station

    2012-07-15

    Make-up water addition to the steam/water cycle at Koeberg Nuclear Power Station usually results in a corresponding increase of the chloride concentration in the steam generator blowdown system. During plant transients, when higher than normal make-up is required to the secondary plant, the concentration of chloride occasionally exceeds the limiting value for the station chemistry performance indicator. Irrespective of this, the demineralised water make-up supply tanks, which are routinely analysed for chloride, are within all recognised acceptable standards for secondary water make-up and therefore these tanks do not initially appear to be the source of chloride contamination. Water treatment at the plant relies essentially on ion exchange, which has been proven to be very effective in removing inorganic ionic species such as chloride. Organic compounds are less effectively removed by ion exchange and may pass through the treatment system, and these organics can reside undetected in the make-up water tanks. Historically, the elevated chloride concentration following high system make-up has been attributed to chlorinated organic compounds known as trihalomethanes being present in the make-up water tanks, but no rigorous study had been undertaken. As it has been assumed that the majority of chloride in the secondary system originates from the make-up water organic impurities, it was considered important to confirm this by compiling an organic profile of the secondary plant. The use of organic additives was also taken into account in the profile. This work has confirmed the contribution from trihalomethanes and has also found that other organochlorides contribute even more significantly to the overall chloride inventory of the secondary plant. (orig.)

  2. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  3. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  4. Energy fluxes in oil palm plantations as affected by water storage in the trunk

    Science.gov (United States)

    Meijide, Ana; Röll, Alexander; Fan, Yuanchao; Herbst, Mathias; Niu, Furong; Tiedemann, Frank; June, Tania; Rauf, Abdul; Hölscher, Dirk; Knohl, Alexander

    2017-04-01

    Oil palm is increasingly expanding, particularly in Indonesia, but information on water and energy fluxes in oil palm plantations is still very limited and on how those are affected by environmental conditions or oil palm age. Using the eddy covariance technique, we studied turbulent fluxes of sensible (H) and latent (LE) heat and gross primary production (GPP) for 8 months each in a young oil palm plantation (1-year old) and subsequently in a mature plantation (12-year old) in Jambi Province, Sumatra, Indonesia. We measured transpiration (T) simultaneously using a sap flux technique. The energy budget was dominated by LE in both plantations, particularly in the mature one, where it represented up to 70% of the available energy. In the young oil palm plantation, evapotranspiration (ET) was significantly reduced and H fluxes were higher. This affected the Bowen ratio, defined as the ratio of H and LE, which was higher in the 1-year old plantation (0.67±0.33), where it remained constant during the day, than in the mature plantation (0.14±0.09), where it varied considerably over the day, suggesting that water accumulated inside the canopy. Using the Community Land Model (CLM), a process based land surface model that has been adapted to oil palm functional traits (i.e. CLM-Palm), we investigated the contribution of different water sources to the measured fluxes. CLM-Palm differentiates leaf and stem surfaces in modelling water interception and is therefore able to diagnose the fraction of dry leaves that contribute to T and the wet fraction of all vegetation surfaces (leaf and stem) that contributes to evaporation. Results from our simulations strengthen our hypothesis of significant contribution of canopy evaporation to ET. As observed in the field, water accumulates inside the canopy in the mature plantation in oil palm trunk surfaces including epiphytes, creating water reservoirs in the trunk, which potentially contribute to ET when they evaporate. The decoupling

  5. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    Energy Technology Data Exchange (ETDEWEB)

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  6. The D&D of the Experimental Boiling Water Reactor (EBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Boling, L.E.; Yule, T.J.; Bhattacharyya, S.K.

    1996-03-01

    Argonne National Laboratory has completed the D&D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D&D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort.

  7. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  8. Variability of carbon and water fluxes following climate extremes over a tropical forest in southwestern Amazonia.

    Directory of Open Access Journals (Sweden)

    Marcelo Zeri

    Full Text Available The carbon and water cycles for a southwestern Amazonian forest site were investigated using the longest time series of fluxes of CO2 and water vapor ever reported for this site. The period from 2004 to 2010 included two severe droughts (2005 and 2010 and a flooding year (2009. The effects of such climate extremes were detected in annual sums of fluxes as well as in other components of the carbon and water cycles, such as gross primary production and water use efficiency. Gap-filling and flux-partitioning were applied in order to fill gaps due to missing data, and errors analysis made it possible to infer the uncertainty on the carbon balance. Overall, the site was found to have a net carbon uptake of ≈5 t C ha(-1 year(-1, but the effects of the drought of 2005 were still noticed in 2006, when the climate disturbance caused the site to become a net source of carbon to the atmosphere. Different regions of the Amazon forest might respond differently to climate extremes due to differences in dry season length, annual precipitation, species compositions, albedo and soil type. Longer time series of fluxes measured over several locations are required to better characterize the effects of climate anomalies on the carbon and water balances for the whole Amazon region. Such valuable datasets can also be used to calibrate biogeochemical models and infer on future scenarios of the Amazon forest carbon balance under the influence of climate change.

  9. Performance of a Cross-Flow Humidifier with a High Flux Water Vapor Transport Membrane

    Energy Technology Data Exchange (ETDEWEB)

    Ahluwalia, R. K.; Wang, X.; Johnson, W. B.; Berg, F.; Kadylak, D.

    2015-09-30

    Water vapor transport (WVT) flux across a composite membrane that consists of a very thin perfluorosulfonic acid (PFSA) ionomer layer sandwiched between two expanded polytetrafluoroethylene (PTFE) microporous layers is investigated. Static and dynamic tests are conducted to measure WVT flux for different composite structures; a transport model shows that the underlying individual resistances for water diffusion in the gas phase and microporous and ionomer layers and for interfacial kinetics of water uptake at the ionomer surface are equally important under different conditions. A finite-difference model is formulated to determine water transport in a full-scale (2-m2 active membrane area) planar cross-flow humidifier module assembled using pleats of the optimized composite membrane. In agreement with the experimental data, the modeled WVT flux in the module increases at higher inlet relative humidity (RH) of the wet stream and at lower pressures, but the mass transfer effectiveness is higher at higher pressures. The model indicates that the WVT flux is highest under conditions that maintain the wet stream at close to 100% RH while preventing the dry stream from becoming saturated. The overall water transport is determined by the gradient in RH of the wet and dry streams but is also affected by vapor diffusion in the gas layer and the microporous layer.

  10. Water Use Patterns of Four Tropical Bamboo Species Assessed with Sap Flux Measurements.

    Science.gov (United States)

    Mei, Tingting; Fang, Dongming; Röll, Alexander; Niu, Furong; Hendrayanto; Hölscher, Dirk

    2015-01-01

    Bamboos are grasses (Poaceae) that are widespread in tropical and subtropical regions. We aimed at exploring water use patterns of four tropical bamboo species (Bambusa vulgaris, Dendrocalamus asper, Gigantochloa atroviolacea, and G. apus) with sap flux measurement techniques. Our approach included three experimental steps: (1) a pot experiment with a comparison of thermal dissipation probes (TDPs), the stem heat balance (SHB) method and gravimetric readings using potted B. vulgaris culms, (2) an in situ calibration of TDPs with the SHB method for the four bamboo species, and (3) field monitoring of sap flux of the four bamboo species along with three tropical tree species (Gmelina arborea, Shorea leprosula, and Hevea brasiliensis) during a dry and a wet period. In the pot experiment, it was confirmed that the SHB method is well suited for bamboos but that TDPs need to be calibrated. In situ, species-specific parameters for such calibration formulas were derived. During field monitoring we found that some bamboo species reached high maximum sap flux densities. Across bamboo species, maximal sap flux density increased with decreasing culm diameter. In the diurnal course, sap flux densities in bamboos peaked much earlier than radiation and vapor pressure deficit (VPD), and also much earlier than sap flux densities in trees. There was a pronounced hysteresis between sap flux density and VPD in bamboos, which was less pronounced in trees. Three of the four bamboo species showed reduced sap flux densities at high VPD values during the dry period, which was associated with a decrease in soil moisture content. Possible roles of internal water storage, root pressure and stomatal sensitivity are discussed.

  11. Aquaporin mediated water flux as a target for diuretic development.

    Science.gov (United States)

    Laski, M E; Pressley, T A

    1999-11-01

    Within the past decade an entire family of membrane proteins--aquaporins--which function as transmembrane water channels has been identified; they occur throughout the plant, animal, and bacterial kingdoms. Several family members permit glycerol and urea permeability. Most aquaporins are inhibited by mercury. Constitutively expressed aquaporin 1 is the major permeability channel of the proximal tubule, descending thin limb of the loop of Henle, and it is also found in vasa recta. Aquaporin 2 is expressed in the principal cells of the collecting duct where it shuttles between intracellular vesicles and the apical membrane in response to vasopressin. Aquaporin 2 mutations cause nephrogenic diabetes insipidus; increased aquaporin 2 activity is implicated in the pathophysiology of heart failure, cirrhosis, and nephrotic syndrome. Aquaporins 3 and 4 provide basolateral membrane water channels in the collecting duct. These 4 channels and 6 others are also found elsewhere throughout the body. The physiological importance of several of the channels remains unknown. Aquaporin 1 inhibitors might induce useful diuresis, but humans who lack aquaporin 1 have no significant clinical disease. Inhibition of aquaporin 2 activity by vasopressin receptor antagonists may be useful in heart failure, cirrhosis, nephrotic syndrome, and the syndrome of inappropriate antidiuretic hormone (ADH) release.

  12. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    Science.gov (United States)

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

  13. Supercritical Water Gasification of Biomass in a Ceramic Reactor: Long-Time Batch Experiments

    Directory of Open Access Journals (Sweden)

    Daniele Castello

    2017-10-01

    Full Text Available Supercritical water gasification (SCWG is an emerging technology for the valorization of (wet biomass into a valuable fuel gas composed of hydrogen and/or methane. The harsh temperature and pressure conditions involved in SCWG (T > 375 °C, p > 22 MPa are definitely a challenge for the manufacturing of the reactors. Metal surfaces are indeed subject to corrosion under hydrothermal conditions, and expensive special alloys are needed to overcome such drawbacks. A ceramic reactor could be a potential solution to this issue. Finding a suitable material is, however, complex because the catalytic effect of the material can influence the gas yield and composition. In this work, a research reactor featuring an internal alumina inlay was utilized to conduct long-time (16 h batch tests with real biomasses and model compounds. The same experiments were also conducted in batch reactors made of stainless steel and Inconel 625. The results show that the three devices have similar performance patterns in terms of gas production, although in the ceramic reactor higher yields of C2+ hydrocarbons were obtained. The SEM observation of the reacted alumina surface revealed a good resistance of such material to supercritical conditions, even though some intergranular corrosion was observed.

  14. Modeling of water lighting process and calculation of the reactor-clarifier to improve energy efficiency

    Science.gov (United States)

    Skolubovich, Yuriy; Skolubovich, Aleksandr; Voitov, Evgeniy; Soppa, Mikhail; Chirkunov, Yuriy

    2017-10-01

    The article considers the current questions of technological modeling and calculation of the new facility for cleaning natural waters, the clarifier reactor for the optimal operating mode, which was developed in Novosibirsk State University of Architecture and Civil Engineering (SibSTRIN). A calculation technique based on well-known dependences of hydraulics is presented. A calculation example of a structure on experimental data is considered. The maximum possible rate of ascending flow of purified water was determined, based on the 24 hour clarification cycle. The fractional composition of the contact mass was determined with minimal expansion of contact mass layer, which ensured the elimination of stagnant zones. The clarification cycle duration was clarified by the parameters of technological modeling by recalculating maximum possible upward flow rate of clarified water. The thickness of the contact mass layer was determined. Likewise, clarification reactors can be calculated for any other lightening conditions.

  15. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

  16. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson, J. Boise; Reid, Robert S.

    2006-01-01

    As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

  17. Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson J. Boise; Reid, Robert S.

    2007-01-01

    As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

  18. Biological mine water treatment operating a one stage reactor system

    CSIR Research Space (South Africa)

    Baloyi, MJ

    2006-05-01

    Full Text Available cuttings, the source of cellulose. The total experimental period was 113 days, which was divided over 4 periods resulting from the addition of fresh grass cuttings and the feed water flow rate. It was concluded from this study that the microorganisms from...

  19. Environmentally assisted cracking in Light Water Reactors. Volume 16: Semiannual report, October 1992--March 1993

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  20. Carbon dioxide and water fluxes in grasslands of Inner Mongolia (China)

    Science.gov (United States)

    Vetter, S.; Ketzer, B.; Grünwald, T.; Bernhofer, Ch.

    2009-04-01

    Grasslands are one of the dominating vegetation types in the world. In China grasslands capture 400 Mha. This huge area has great influence on water and carbon stocks and fluxes. Water and carbon exchange influence the local concentration of greenhouse gases. In the steppe of Inner Mongolia there are problems of overgrazing, erosion and ongoing desertification. Through these processes the seasonal patterns of the water and carbon cycles are changed. Within the project MAGIM (Matter fluxes in grasslands of Inner Mongolia as influenced by stocking rate), which is a multidisciplinary project bringing together German and Chinese expertise, matter fluxes were measured with eddy covariance method. In this study the results of MAGIM concerning the carbon dioxide and water fluxes were presented. The study site is in the Xilin River catchment in the Northeast of Inner Mongolia Autonomous Region, China. The region is a continental temperate semiarid zone with cold dry winters and warm humid summers. The annual mean temperature is about 2 °C and the annual precipitation is 350 mm. The EC-measurements include measurements at different steppe types (Leymus chinensis, Stipa grandis) and various land use (overgrazed, winter grazed, continuously crazed, ungrazed since 1979) by one permanent and by two roving towers. From 2002 to 2004 there were continuous measurements at the ungrazed site (Leymus chinensis). The roving tower was used in the vegetation period at the other sites. From 2005 a third tower was available to measure at the overgrazed site continuously. The results show large differences in the carbon dioxide fluxes of the various land use. In general the carbon dioxide fluxes are small in Inner Mongolia. At the ungrazed site the results show negative net carbon exchange (CO2 sink). The positive net carbon exchange at overgrazed site indicates a CO2 source. The partially grazed sites switch between CO2 sink and source dependent on the environmental conditions. Lower

  1. Assessing HYDRUS-2D model to estimate soil water contents and olive tree transpiration fluxes under different water distribution systems

    Science.gov (United States)

    Autovino, Dario; Negm, Amro; Rallo, Giovanni; Provenzano, Giuseppe

    2016-04-01

    In Mediterranean countries characterized by limited water resources for agricultural and societal sectors, irrigation management plays a major role to improve water use efficiency at farm scale, mainly where irrigation systems are correctly designed to guarantee a suitable application efficiency and the uniform water distribution throughout the field. In the last two decades, physically-based agro-hydrological models have been developed to simulate mass and energy exchange processes in the soil-plant-atmosphere (SPA) system. Mechanistic models like HYDRUS 2D/3D (Šimunek et al., 2011) have been proposed to simulate all the components of water balance, including actual crop transpiration fluxes estimated according to a soil potential-dependent sink term. Even though the suitability of these models to simulate the temporal dynamics of soil and crop water status has been reported in the literature for different horticultural crops, a few researches have been considering arboreal crops where the higher gradients of root water uptake are the combination between the localized irrigation supply and the three dimensional root system distribution. The main objective of the paper was to assess the performance of HYDRUS-2D model to evaluate soil water contents and transpiration fluxes of an olive orchard irrigated with two different water distribution systems. Experiments were carried out in Castelvetrano (Sicily) during irrigation seasons 2011 and 2012, in a commercial farm specialized in the production of table olives (Olea europaea L., var. Nocellara del Belice), representing the typical variety of the surrounding area. During the first season, irrigation water was provided by a single lateral placed along the plant row with four emitters per plant (ordinary irrigation), whereas during the second season a grid of emitters laid on the soil was installed in order to irrigate the whole soil surface around the selected trees. The model performance was assessed based on the

  2. Cross-scale impact of climate temporal variability on ecosystem water and carbon fluxes

    Science.gov (United States)

    Paschalis, Athanasios; Fatichi, Simone; Katul, Gabriel G.; Ivanov, Valeriy Y.

    2015-09-01

    While the importance of ecosystem functioning is undisputed in the context of climate change and Earth system modeling, the role of short-scale temporal variability of hydrometeorological forcing (~1 h) on the related ecosystem processes remains to be fully understood. Various impacts of meteorological forcing variability on water and carbon fluxes across a range of scales are explored here using numerical simulations. Synthetic meteorological drivers that highlight dynamic features of the short temporal scale in series of precipitation, temperature, and radiation are constructed. These drivers force a mechanistic ecohydrological model that propagates information content into the dynamics of water and carbon fluxes for an ensemble of representative ecosystems. The focus of the analysis is on a cross-scale effect of the short-scale forcing variability on the modeled evapotranspiration and ecosystem carbon assimilation. Interannual variability of water and carbon fluxes is emphasized in the analysis. The main study inferences are summarized as follows: (a) short-scale variability of meteorological input does affect water and carbon fluxes across a wide range of time scales, spanning from the hourly to the annual and longer scales; (b) different ecosystems respond to the various characteristics of the short-scale variability of the climate forcing in various ways, depending on dominant factors limiting system productivity; (c) whenever short-scale variability of meteorological forcing influences primarily fast processes such as photosynthesis, its impact on the slow-scale variability of water and carbon fluxes is small; and (d) whenever short-scale variability of the meteorological forcing impacts slow processes such as movement and storage of water in the soil, the effects of the variability can propagate to annual and longer time scales.

  3. Behaviour of carbon dioxide and water vapour flux densities from a disturbed raised peat bog

    NARCIS (Netherlands)

    Nieveen, J.P.; Jacobs, A.F.G.

    2002-01-01

    Measurements of carbon dioxide and water vapour flux densities were carried out for a disturbed raised peat bog in the north of the Netherlands during an 18 month continuous experiment. Tussock grass (sp. Molinea caerulae) mainly dominated the vegetation of the bog area. The maximum leaf area index

  4. Energy and water cycle over the Tibetan plateau : surface energy balance and turbulent heat fluxes

    NARCIS (Netherlands)

    Su, Zhongbo; Zhang, Ting; Ma, Yaoming; Jia, Li; Wen, Jun

    2006-01-01

    This contribution presents an overview and an outlook of studies on energy and water cycle over the Tibetan plateau with focuses on the estimation of energy balance terms and turbulent heat fluxes. On the basis of the surface energy balance calculations, we show that the phenomena of the energy

  5. Energy and water cycle over the Tibetan Plateau: surface energy balance and turbulent heat fluxes

    NARCIS (Netherlands)

    Su, Z.; Zhang, T.; Ma, Y.; Jia, L.; Wen, J.

    2006-01-01

    This contribution presents an overview and an outlook of studies on energy and water cycle over the Tibetan plateau with focuses on the estimation of energy balance terms and turbulent heat fluxes. On the basis of the surface energy balance calculations, we show that the phenomena of the energy

  6. A new automated passive capillary lysimeter for logging real-time drainage water fluxes

    Science.gov (United States)

    Effective monitoring of chemical transport through the soil profile requires accurate and appropriate instrumentation to measure drainage water fluxes below the root zone of cropping system. The objectives of this study were to methodically describe in detail the construction and installation of a n...

  7. Residual fluxes of salt and water in the Azhikode estuary, west coast of India

    Digital Repository Service at National Institute of Oceanography (India)

    Pylee, A.; Varma, P.U.; Revichandran, C.

    hours at all stations and the data were analysed to provide estimates of the residual fluxes of water and salt. The interpolated data for the non-dimensional depth was used for computation of depth, tide and cross sectional averages. A net seaward flow...

  8. Analysis of carbon dioxide, water vapour and energy fluxes over an ...

    Indian Academy of Sciences (India)

    temperate grassland ecosystem; Bound.-Layer Meteorol. 52 135–149. Mahrt L and Dean V 2002 Relationship of area-averaged carbon dioxide and water vapour fluxes to atmospheric variables; Agric. For. Meteor. 112 195–202. Ohtaki E 1984 Application of an infrared carbon dioxide and humidity instrument to studies of ...

  9. Study of neutronic flux in IPR-R1 reactor with MCNPX; Estudo do fluxo neutronico no reator IPR-R1 com o MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Melo, J.A.S.; Castrillo, L.S., E-mail: julio.angelo@poli.br, E-mail: lazara@poli.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica; Oliveira, R.M.B.M., E-mail: romero.matias@educacao.pe.gov.br [Secretaria Executiva de Educacao do Estado de Pernambuco (SEE), Recife, PE (Brazil)

    2016-11-01

    MCNPX computer code, one of the latest versions of code MCNP transport were used to study the flux distribution and its neutronic fluence as a function of energy in two research reactor irradiation IPR-R1. The model developed was validated with research conducted by Dalle (2005). Initially, in the simulation is considered fresh fuel whose core configuration contained three neutron rods control, being two of them 100% ejected while the other inserted 3,1 x 10{sup -1} m deep, as adopted in the literature situation. The neutron source used was the critical type, through KSRC card. The results of the neutron flow and neutronic fluence were obtained in the central tube and the turntable on a range of energy spectrum that ranged from 1.0 x 10{sup -9} MeV to 10 MeV, showing good correlations with the model used in validation. Finally, a hypothetical situation wherein the three reactor control rods are ejected simultaneously was simulated. The simulation results showed an increase in the neutron flux of 7% in the central tube and 5% on the turntable.

  10. Assessment of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    OpenAIRE

    Zhang, Z.; Dong, Y; Scherer, W

    2005-01-01

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemente...

  11. Implications for accident management of adding water to a degrading reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-02-01

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

  12. Model reactor for photocatalytic degradation of persistent chemicals in ponds and waste water.

    Science.gov (United States)

    Franke, R; Franke, C

    1999-12-01

    A laboratory scale flow-through model reactor for the degradation of persistent chemicals using titanium dioxide (TiO2) as photocatalyst immobilized on glass beads is presented. In the test system with a volume of 18 L contaminated water is pumped to the upper part of the floating reactor and flows over the coated beads which are exposed to UV-radiation. The degradation of two dyes of different persistence was investigated. Primary degradation of methylene blue did not fit a first order kinetic due to coincident adsorption onto the photocatalyst and direct photolysis, resulting in a half-life of 6 h. A filtrate of a green algae suspension accelerated the colour removal. In contrast, reactive red 2 was degraded only by photocatalysis; neither adsorption nor direct photolysis led to a colour removal. The course of primary degradation followed a first order kinetic with a half-life of 18 h and a rate constant of 0.04 h-1. Analysis of the degradation products indicated mineralization by detection of NO2- and NO3-, accompanied by a decrease of pH and an increase of conductivity. A successful adaptation of the model reactor (scale 1:10) to dimensions required for surface waters and waste water treatment plants would be a cost-efficient and environmentally sustainable application of photocatalysis for the treatment of industrially polluted water and could be of relevance for third world countries, particularly those favoured by high solar radiation.

  13. Boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Tokarev, Yu.I.; Sokolov, I.N.; Skvortsov, S.A.; Sidorov, A.M.; Krauze, L.V.

    1978-04-01

    The possibility of using a boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant (CHPP) was considered, with design features of the reactor intended for a two-purpose plant. A prestressed reinforced concrete vessel and integral arrangement of the primary circuit ensured reliability of the atomic CHPP using various CHPP flowsheets.

  14. Estimating greenhouse gas fluxes from constructed wetlands used for water quality improvement

    Directory of Open Access Journals (Sweden)

    Sukanda Chuersuwan

    2014-06-01

    Full Text Available Methane (CH4 , nitrous oxide (N2O and carbon dioxide (CO2 fluxes were evaluated from constructed wetlands (CWs used to improve domestic wastewater quality. Experiments employed subsurface flow (SF and free water surface flow (FWS CWs planted with Cyperus spp. Results showed seasonal fluctuations of greenhouse gas fluxes. Greenhouse gas fluxes from SF-CWs and FWS-CWS were significantly different (p<0.05 while pollutant removal efficiencies of both CWs were not significantly different. The average CH4 , N2O and CO2 fluxes from SF-CWs were 2.9±3.5, 1.0±1.7, and 15.2±12.3 mg/m2 /hr, respectively, corresponding to the average global warming potential (GWP of 392 mg CO2 equivalents/m2 /hr. For FWS-CWs, the average CH4 , N2O and CO2 fluxes were 5.9±4.8, 1.8±1.0, and 29.6±20.2 mg/m2 /hr, respectively, having an average GWP of 698 mg CO2 equivalents/m2 /hr. Thus, FWS-CWs have a higher GWP than SF-CWs when they were used as a system for domestic water improvement.

  15. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  16. Seasonal changes in soil water repellency and their effect on soil CO2 fluxes

    Science.gov (United States)

    Urbanek, Emilia; Qassem, Khalid

    2016-04-01

    Soil water repellency (SWR) is a seasonally variable phenomenon controlled by moisture content and at the same time a regulator of the distribution and conductivity of water in the soil. The distribution and availability of water in soil is also an important factor for microbial activity, decomposition of soil organic matter and exchange of gases like CO2 and CH4 between the soil and the atmosphere. It has been therefore hypothesised that SWR by restricting water availability in soil can affect the production and the transport of CO2 in the soil and between the soil and the atmosphere. This study investigates the effect of seasonal changes in soil moisture and water repellency on CO2 fluxes from soil. The study was conducted for 3 year at four grassland and pine forest sites in the UK with contrasting precipitation. The results show the temporal changes in soil moisture content and SWR are affected by rainfall intensity and the length of dry periods between the storms. Soils exposed to very high annual rainfall (>1200mm) can still exhibit high levels of SWR for relatively long periods of time. The spatial variation in soil moisture resulting from SWR affects soil CO2 fluxes, but the most profound effect is visible during and immediately after the rainfall events. Keywords: soil water repellency, CO2 flux, hydrophobicity, preferential flow, gas exchange, rainfall

  17. Pretreatment of agriculture field water for improving membrane flux during pesticide removal

    Science.gov (United States)

    Mehta, Romil; Saha, N. K.; Bhattacharya, A.

    2017-10-01

    Pretreatment of feed water to improve membrane flux during filtration of agriculture field water containing substituted phenyl urea pesticide diuron has been reported. Laboratory-made reverse osmosis membrane was used for filtration. Preliminary experiments were conducted with model solution containing natural organic matter extracted from commercial humic acids, divalent ions Ca2+, Mg2+. Membrane fouling was characterized by pure water flux decline, change in membrane hydrophilicity and infrared spectroscopy. Natural organic matter present in field water causes severe membrane fouling. The presence of divalent cations further aggravated fouling. Use of ethylenediaminetetraacetic acid (EDTA) and polyacrylic acids (PAA) in feed resulted in the decrease in membrane fouling. Pretreatment of field water is a must if it is contaminated with micro-organism having membrane fouling potential. Feed water pretreatment and use of PAA restricted membrane fouling to 16 % after 60 h of filtration. Membrane permeate flux decline was maximum at the first 12 h and thereafter remained steady at around 45-46 lm-2h-1 till the end of 60 h. Diuron rejection remained consistently greater than 93 % throughout the experiment. Diuron rejection was found to be unaffected by membrane fouling.

  18. Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mohora, Emilijan, E-mail: emohora@ifc.org [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia); Roncevic, Srdjan; Dalmacija, Bozo; Agbaba, Jasmina; Watson, Malcolm; Karlovic, Elvira; Dalmacija, Milena [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A continuous electrocoagulation/flotation reactor was designed built and operated. Black-Right-Pointing-Pointer Highest NOM removal according to UV{sub 254} was 77% relative to raw groundwater. Black-Right-Pointing-Pointer Highest NOM removal accordance to DOC was 71%, relative to raw groundwater. Black-Right-Pointing-Pointer Highest As removal archived was 85% (6.2 {mu}g/l), relative to raw groundwater. Black-Right-Pointing-Pointer Specific reactor energy and electrode consumption was 1.7 kWh/m{sup 3} and 66 g Al/m{sup 3}. - Abstract: The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate = 4.3 l/h, inter electrode distance = 2.8 cm, current density = 5.78 mA/cm{sup 2}, A/V ratio = 0.248 cm{sup -1}. The NOM removal according to UV{sub 254} absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 {mu}g As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m{sup 3}. According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater.

  19. The water vapour flux above Switzerland and its role in the August 2005 extreme precipitation and flooding

    Energy Technology Data Exchange (ETDEWEB)

    N' Dri Koffi, Ernest; Maetzler, Christian [Bern Univ. (Switzerland). Inst. of Applied Physics; Graham, Edward [Bern Univ. (Switzerland). Inst. of Applied Physics; University of the Highlands and Islands, Stornoway, Scotland (United Kingdom). Lews Castle College

    2013-10-15

    The water budget approach is applied to an atmospheric box above Switzerland (hereafter referred to as the 'Swiss box') to quantify the atmospheric water vapour flux using ECMWF ERA-Interim reanalyses. The results confirm that the water vapour flux through the Swiss box is highly temporally variable, ranging from 1 to 5 x 10{sup 7} kg/s during settled anticyclonic weather, but increasing in size by a factor of ten or more during high speed currents of water vapour. Overall, Switzerland and the Swiss box 'import' more water vapour than it 'exports', but the amount gained remains only a small fraction (1% to 5%) of the total available water vapour passing by. High inward water vapour fluxes are not necessarily linked to high precipitation episodes. The water vapour flux during the August 2005 floods, which caused severe damage in central Switzerland, is examined and an assessment is made of the computed water vapour fluxes compared to high spatio-temporal rain gauge and radar observations. About 25% of the incoming water vapour flux was stored in Switzerland. The computed water vapour fluxes from ECMWF data compare well with the mean rain gauge observations and the combined rain-gauge radar precipitation products. (orig.)

  20. The water vapour flux above Switzerland and its role in the August 2005 extreme precipitation and flooding

    Directory of Open Access Journals (Sweden)

    Ernest N'dri Koffi

    2013-07-01

    Full Text Available The water budget approach is applied to an atmospheric box above Switzerland (hereafter referred to as the “Swiss box” to quantify the atmospheric water vapour flux using ECMWF ERA-Interim reanalyses. The results confirm that the water vapour flux through the Swiss box is highly temporally variable, ranging from 1 to 5 · 107 kg/s during settled anticyclonic weather, but increasing in size by a factor of ten or more during high speed currents of water vapour. Overall, Switzerland and the Swiss box “import” more water vapour than it “exports”, but the amount gained remains only a small fraction (1% to 5% of the total available water vapour passing by. High inward water vapour fluxes are not necessarily linked to high precipitation episodes. The water vapour flux during the August 2005 floods, which caused severe damage in central Switzerland, is examined and an assessment is made of the computed water vapour fluxes compared to high spatio-temporal rain gauge and radar observations. About 25% of the incoming water vapour flux was stored in Switzerland. The computed water vapour fluxes from ECMWF data compare well with the mean rain gauge observations and the combined rain-gauge radar precipitation products.

  1. Generalized analytical model for benthic water flux forced by surface gravity waves

    Science.gov (United States)

    King, J.N.; Mehta, A.J.; Dean, R.G.

    2009-01-01

    A generalized analytical model for benthic water flux forced by linear surface gravity waves over a series of layered hydrogeologic units is developed by adapting a previous solution for a hydrogeologic unit with an infinite thickness (Case I) to a unit with a finite thickness (Case II) and to a dual-unit system (Case III). The model compares favorably with laboratory observations. The amplitude of wave-forced benthic water flux is shown to be directly proportional to the amplitude of the wave, the permeability of the hydrogeologic unit, and the wave number and inversely proportional to the kinematic viscosity of water. A dimensionless amplitude parameter is introduced and shown to reach a maximum where the product of water depth and the wave number is 1.2. Submarine groundwater discharge (SGD) is a benthic water discharge flux to a marine water body. The Case I model estimates an 11.5-cm/d SGD forced by a wave with a 1 s period and 5-cm amplitude in water that is 0.5-m deep. As this wave propagates into a region with a 0.3-m-thick hydrogeologic unit, with a no-flow bottom boundary, the Case II model estimates a 9.7-cm/d wave-forced SGD. As this wave propagates into a region with a 0.2-m-thick hydrogeologic unit over an infinitely thick, more permeable unit, the Case III quasi-confined model estimates a 15.7-cm/d wave-forced SGD. The quasi-confined model has benthic constituent flux implications in coral reef, karst, and clastic regions. Waves may undermine tracer and seepage meter estimates of SGD at some locations. Copyright 2009 by the American Geophysical Union.

  2. A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

    Directory of Open Access Journals (Sweden)

    Diego Mandelli

    2015-01-01

    Full Text Available In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

  3. International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

    Science.gov (United States)

    Geraskin, N. I.; Glebov, V. B.

    2017-01-01

    The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

  4. Effects of salt pond restoration on benthic flux: Sediment as a source of nutrients to the water column

    Science.gov (United States)

    Topping, Brent R.; Kuwabara, James S.; Carter, James L.; Garrettt, Krista K.; Mruz, Eric; Piotter, Sarah; Takekawa, John Y.

    2016-01-01

    Understanding nutrient flux between the benthos and the overlying water (benthic flux) is critical to restoration of water quality and biological resources because it can represent a major source of nutrients to the water column. Extensive water management commenced in the San Francisco Bay, Beginning around 1850, San Francisco Bay wetlands were converted to salt ponds and mined extensively for more than a century. Long-term (decadal) salt pond restoration efforts began in 2003. A patented device for sampling porewater at varying depths, to calculate the gradient, was employed between 2010 and 2012. Within the former ponds, the benthic flux of soluble reactive phosphorus and that of dissolved ammonia were consistently positive (i.e., moving out of the sediment into the water column). The lack of measurable nitrate or nitrite concentration gradients across the sediment-water interface suggested negligible fluxes for dissolved nitrate and nitrite. The dominance of ammonia in the porewater indicated anoxic sediment conditions, even at only 1 cm depth, which is consistent with the observed, elevated sediment oxygen demand. Nearby openestuary sediments showed much lower benthic flux values for nutrients than the salt ponds under resortation. Allochthonous solute transport provides a nutrient advective flux for comparison to benthic flux. For ammonia, averaged for all sites and dates, benthic flux was about 80,000 kg/year, well above the advective flux range of −50 to 1500 kg/year, with much of the variability depending on the tidal cycle. By contrast, the average benthic flux of soluble reactive phosphorus was about 12,000 kg/year, of significant magnitude, but less than the advective flux range of 21,500 to 30,000 kg/year. These benthic flux estimates, based on solute diffusion across the sediment-water interface, reveal a significant nutrient source to the water column of the pond which stimulates algal blooms (often autotrophic). This benthic source may be

  5. Mixing of cooling water in the mixing chambers of the HPLWR-High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wank, Alexander, E-mail: alexander.wank@siemens.co [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies, P.O. Box 3640, 76021 Karlsruhe (Germany); Starflinger, Joerg; Schulenberg, Thomas [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies, P.O. Box 3640, 76021 Karlsruhe (Germany); Laurien, Eckart [University of Stuttgart, Institute for Nuclear Technology and Energy Systems (IKE) Pfaffenwaldring 31, D-70550 Stuttgart (Germany)

    2010-10-15

    The High Performance Light Water Reactor (HPLWR), a supercritical water cooled reactor concept with multiple heat-up steps, requires efficient mixing of the coolant between these steps to minimize hot spots in the core. Analyzing and improving the mixing in the mixing chamber above the core, situated between evaporator and superheater assemblies, and below the core, between the first and second superheater, is one of the challenges in the design process of the HPLWR. Different measures to enhance mixing have been studied with CFD analyses, in which a new design approach has been applied to the upper mixing chamber. It simplifies the complex structures and takes the effects of the disregarded structures into account by introducing source terms into the momentum equations.

  6. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  7. Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

    Energy Technology Data Exchange (ETDEWEB)

    Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

    2011-08-01

    Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

  8. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  9. Light-water-reactor safety research program: quarterly progress report, July--September 1977

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-01-01

    Progress is summarized on the Argonne National Laboratory work performed during July, August, and September 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of Zircaloy containing oxygen; and (4) steam-explosion studies.

  10. 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.

  11. Analysis of a Partial MOX Core Design with Tritium Targets for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anistratov, Dmitriy Y. [Texas A & M Univ., College Station, TX (United States); Adams, Marvin L. [Texas A & M Univ., College Station, TX (United States)

    1998-04-19

    This report constitutes tangible and verifiable deliverable associated with the task To study the effects of using WG MOX fuel in tritium-producing LWR” of the subproject Water Reactor Options for Disposition of Plutonium. The principal investigators of this subproject are Naeem M. Abdurrahman of the University of Texas at Austin and Marvin L. Adams of Texas A&M University. This work was sponsored by the Amarillo National Resource Center for Plutonium.

  12. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

    Energy Technology Data Exchange (ETDEWEB)

    Essadki, A.H., E-mail: essadki@hotmail.com [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Gourich, B. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Vial, Ch. [Laboratoire de Genie Chimique et Biochimique, LGCB-UBP/ENSCCF, 24 avenue des Landais, BP 206, 63174 Aubiere Cedex (France); Delmas, H. [Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France); Bennajah, M. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France)

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm{sup 2}, but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H{sub 2} microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  13. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor.

    Science.gov (United States)

    Essadki, A H; Gourich, B; Vial, Ch; Delmas, H; Bennajah, M

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12mA/cm(2), but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H(2) microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  14. Filtration behavior of casein glycomacropeptide (CGMP) in an enzymatic membrane reactor: fouling control by membrane selection and threshold flux operation

    DEFF Research Database (Denmark)

    Luo, Jianquan; Morthensen, Sofie Thage; Meyer, Anne S.

    2014-01-01

    Sialylated human milk oligosaccharides (HMOs) can be produced by enzymatic trans-sialidation using casein glycomacropeptide (CGMP) as the substrate. By performing the reaction in an enzymatic membrane reactor (EMR), simultaneous separation of the HMOs from CGMP and enzyme reuse can be achieved...

  15. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  16. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Science.gov (United States)

    2010-01-01

    ...-Cooled Power Reactors J Appendix J to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. J Appendix J to Part 50—Primary Reactor Containment... basis accident and specified either in the technical specification or associated bases. J. Pt (p.s.i.g...

  17. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    Bragg-Sitton, Shannon M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  18. IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, K. [Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria); Aksan, S. N. [International Atomic Energy Agency, 1400 Vienna (Austria)

    2012-07-01

    The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

  19. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment.

  20. Investigating spatial variability of vertical water fluxes through the streambed in distinctive stream morphologies using temperature and head data

    Science.gov (United States)

    Wang, Liping; Jiang, Weiwei; Song, Jinxi; Dou, Xinyi; Guo, Hongtao; Xu, Shaofeng; Zhang, Guotao; Wen, Ming; Long, Yongqing; Li, Qi

    2017-08-01

    Investigating the interaction of groundwater and surface water is key to understanding the hyporheic processes. The vertical water fluxes through a streambed were determined using Darcian flux calculations and vertical sediment temperature profiles to assess the pattern and magnitude of groundwater/surface-water interaction in Beiluo River, China. Field measurements were taken in January 2015 at three different stream morphologies including a meander bend, an anabranching channel and a straight stream channel. Despite the differences of flux direction and magnitude, flux directions based on vertical temperature profiles are in good agreement with results from Darcian flux calculations at the anabranching channel, and the Kruskal-Wallis tests show no significant differences between the estimated upward fluxes based on the two methods at each site. Also, the upward fluxes based on the two methods show similar spatial distributions on the streambed, indicating (1) that higher water fluxes at the meander bend occur from the center of the channel towards the erosional bank, (2) that water fluxes at the anabranching channel are higher near the erosional bank and in the center of the channel, and (3) that in the straight channel, higher water fluxes appear from the center of the channel towards the depositional bank. It is noted that higher fluxes generally occur at certain locations with higher streambed vertical hydraulic conductivity ( K v) or where a higher vertical hydraulic gradient is observed. Moreover, differences of grain size, induced by stream morphology and contrasting erosional and depositional conditions, have significant effects on streambed K v and water fluxes.

  1. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  2. Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

    Science.gov (United States)

    Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

    2018-01-01

    The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

  3. Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

  4. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  5. Ground-Water Nutrient Flux to Coastal Waters and Numerical Simulation of Wastewater Injection at Kihei, Maui, Hawaii

    Science.gov (United States)

    Hunt, Charles D.

    2007-01-01

    Water sampling and numerical modeling were used to estimate ground-water nutrient fluxes in the Kihei area of Maui, where growth of macroalgae (seaweed) on coral reefs raises ecologic concerns and accumulation on beaches has caused odor and removal problems. Fluxes and model results are highly approximate, first-order estimates because very few wells were sampled and there are few field data to constrain model calibration. Ground-water recharge was estimated to be 22.6 Mgal/d (million gallons per day) within a 73-square-mile area having a coastline length of 8 miles or 13 km (kilometers). Nearly all of the recharge discharges at the coast because ground-water withdrawals are small. Another 3.0 Mgal/d of tertiary-treated wastewater effluent is injected into the regional aquifer at a County treatment plant midway along the coast and about a mile from shore. The injection plume is 0.93 miles wide (1.5 km) at the shore, as estimated from a three-dimensional numerical ground-water model. Wastewater injected beneath the brackish ground-water lens rise