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Sample records for boiling reactor experiment 3

  1. (Boiling water reactor (BWR) CORA experiments)

    SciTech Connect

    Ott, L.J.

    1990-10-16

    To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.

  2. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  3. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  4. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    SciTech Connect

    Wheeler, Timothy A.; Liao, Huafei

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  5. The startup of the Dodewaard natural circulation boiling water reactor -- Experiences

    SciTech Connect

    Nissen, W.H.M.; Van Der Voet, J.; Karuza, J. )

    1994-07-01

    Because of its similarity to the simplified boiling water reactor (SBWR), the Dodewaard natural circulation boiling water reactor (BWR) is of special interest to further development of the SBWR design. It has become especially important to gain more insight into the Dodewaard BWR behavior during startup, paying special attention to its stability. Therefore, special instrumentation was used by means of which a series of measurements were taken during the two startups in February and June 1992. The results obtained from these measurements are used to deepen insight into the recirculation flow and the stability of the reactor during startup under conditions with a normal pressure/power trajectory. They have already shown a very early recirculation flow onset during low-power operation and no indication of reactor instability. Furthermore, they will be used as a basis for the research program investigating the reactor behavior under different pressure/power conditions, which is scheduled for next year.

  6. Microheater Array Boiling Experiment

    NASA Technical Reports Server (NTRS)

    Kim, Jungho; McQuillen, John; Balombin, Joe

    2002-01-01

    By conducting pool boiling tests in microgravity, the effect of buoyancy on the overall boiling process and the relative magnitude of other phenomena can be assessed. Data from KC-135 and sounding rocket experiments indicate little effect of gravity on boiling heat transfer at wall superheats below 25 C, despite vast differences in bubble behavior between gravity levels. In microgravity, a large primary bubble, surrounded by smaller satellite bubbles, moved over the surface, occasionally causing nucleation. Once formed, the primary bubble size remained constant for a given superheat, indicating evaporation at the bubble base is balanced with condensation on the bubble cap. The primary bubble's size increased with wall superheat. Most heaters under the primary bubble had low heat transfer rates, suggesting liquid dryout. Strong Marangoni convection developed in microgravity, forming a 'jet' into the bulk liquid that forced the bubble onto the heater. An experiment is being designed for the. Microgravity Science Glovebox. This experiment uses two 96 element microheater arrays, 2.7 and 7.0 mm in size. These heaters are individually controlled to operate at a constant temperature, measuring local heat fluxes as a function of time and space. Most boiling experiments operate at constant wall heat flux with larger heaters, allowing only time and space-averaged measurements. Each heater is about the bubble departure size in normal gravity, but significantly smaller than the bubble departure size in reduced gravity.

  7. Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications

    NASA Astrophysics Data System (ADS)

    Thomas, Justin W.

    2006-12-01

    The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.

  8. CHIMNEY FOR BOILING WATER REACTOR

    DOEpatents

    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)

  9. SUPERHEATING IN A BOILING WATER REACTOR

    DOEpatents

    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.

  10. Hybrid Reactor Simulation of Boiling Water Reactor Power Oscillations

    SciTech Connect

    Huang Zhengyu; Edwards, Robert M.

    2003-08-15

    Hybrid reactor simulation (HRS) of boiling water reactor (BWR) instabilities, including in-phase and out-of-phase (OOP) oscillations, has been implemented on The Pennsylvania State University TRIGA reactor. The TRIGA reactor's power response is used to simulate reactor neutron dynamics for in-phase oscillation or the fundamental mode of the reactor modal kinetics for OOP oscillations. The reactor power signal drives a real-time boiling channel simulation, and the calculated reactivity feedback is in turn fed into the TRIGA reactor via an experimental changeable reactivity device. The thermal-hydraulic dynamics, together with first harmonic mode power dynamics, is digitally simulated in the real-time environment. The real-time digital simulation of boiling channel thermal hydraulics is performed by solving constitutive equations for different regions in the channel and is realized by a high-performance personal computer. The nonlinearity of the thermal-hydraulic model ensures the capability to simulate the oscillation phenomena, limit cycle and OOP oscillation, in BWR nuclear power plants. By adjusting reactivity feedback gains for both modes, various oscillation combinations can be realized in the experiment. The dynamics of axially lumped power distribution over the core is displayed in three-dimensional graphs. The HRS reactor power response mimics the BWR core-wide power stability phenomena. In the OOP oscillation HRS, the combination of reactor response and the simulated first harmonic power using shaping functions mimics BWR regional power oscillations. With this HRS testbed, a monitoring and/or control system designed for BWR power oscillations can be experimentally tested and verified.

  11. (Severe accident technology of BWR (Boiling Water Reactor) reactors)

    SciTech Connect

    Ott, L.J.

    1989-10-23

    The traveler attended the 1989 CORA Workshop at KfK, FRG. Participation included the presentation included the presentation of three papers on work performed by the Boiling Water Reactor Severe Accident Technology (BWRSAT) program at Oak Ridge National Laboratory (ORNL) in Boiling Water Reactor (BWR) severe accident analyses. The Statement of Work (June 1989) for the BWRSAT Program provides for code analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that BWRSAT personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to arrange for acquisition of detailed CORA facility drawings, experimental data, and related engineering. 17 refs.

  12. Flow Boiling and Condensation Experiment

    NASA Video Gallery

    The Flow Boiling and Condensation Experiment is another investigation that examines the flow of a mixture of liquids and the vapors they produce when in contact with hot space system equipment. Coo...

  13. BOILING SLURRY REACTOR AND METHOD FO CONTROL

    DOEpatents

    Petrick, M.; Marchaterre, J.F.

    1963-05-01

    The control of a boiling slurry nuclear reactor is described. The reactor consists of a vertical tube having an enlarged portion, a steam drum at the top of the vertical tube, and at least one downcomer connecting the steam drum and the bottom of the vertical tube, the reactor being filled with a slurry of fissionabie material in water of such concentration that the enlarged portion of the vertical tube contains a critical mass. The slurry boils in the vertical tube and circulates upwardly therein and downwardly in the downcomer. To control the reactor by controlling the circulation of the slurry, a gas is introduced into the downcomer. (AEC)

  14. New generation of NPP with boiling water reactor of improved safety

    SciTech Connect

    Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.; Mikhan, V.I.; Tokarev, Yu.I.; Cherkashov, Yu.M.; Sokolov, I.N.; Iljin, Yu.V.; Pakh, E.E.; Abramov, V.I.

    1993-12-31

    The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

  15. Superfund record of decision (EPA region 10): Idaho National Engineering Lab, (USDOE) Operable Unit 26 (Stationary Low-Power Reactor-1 and Boiling Water Reactor Experiment-I Burial Grounds), Idaho Falls, ID, December 1, 1995

    SciTech Connect

    1997-03-01

    This document presents the selected remedial action for the Stationary Low-Power Reactor-1 (SL-1) burial ground, the Boiling Water Reactor Experiment-I (BORAX-I) burial ground, and 10 no action sites in Waste Area Group 5. Actual or threatened releases of hazardous substances from the SL-1 and BORAX-I burial grounds, if not addressed by implementing the response action selected in this Record of Decision, may present a current or potential threat to public health, welfare, or the environment. The 10 no action sites do not present a threat to human health or the environment.

  16. Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

    NASA Astrophysics Data System (ADS)

    Seiler, J. M.; Rameau, B.

    Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

  17. Pool Boiling Experiment Has Five Successful Flights

    NASA Technical Reports Server (NTRS)

    Chiaramonte, Fran

    1997-01-01

    The Pool Boiling Experiment (PBE) is designed to improve understanding of the fundamental mechanisms that constitute nucleate pool boiling. Nucleate pool boiling is a process wherein a stagnant pool of liquid is in contact with a surface that can supply heat to the liquid. If the liquid absorbs enough heat, a vapor bubble can be formed. This process occurs when a pot of water boils. On Earth, gravity tends to remove the vapor bubble from the heating surface because it is dominated by buoyant convection. In the orbiting space shuttle, however, buoyant convection has much less of an effect because the forces of gravity are very small. The Pool Boiling Experiment was initiated to provide insight into this nucleate boiling process, which has many earthbound applications in steamgeneration power plants, petroleum plants, and other chemical plants. In addition, by using the test fluid R-113, the Pool Boiling Experiment can provide some basic understanding of the boiling behavior of cryogenic fluids without the large cost of an experiment using an actual cryogen.

  18. Pool Boiling Experiment Has Successful Flights

    NASA Technical Reports Server (NTRS)

    1996-01-01

    The Pool Boiling Experiment (PBE) is designed to improve understanding of the fundamental mechanisms that constitute nucleate pool boiling. Nucleate pool boiling is a process wherein a stagnant pool of liquid is in contact with a surface that can supply heat to the liquid. If the liquid absorbs enough heat, a vapor bubble can be formed. This process occurs when a pot of water boils. On Earth, gravity tends to remove the vapor bubble from the heating surface because it is dominated by buoyant convection. In the orbiting space shuttle, however, buoyant convection has much less of an effect because the forces of gravity are very small. The Pool Boiling Experiment was initiated to provide insight into this nucleate boiling process, which has many Earthbound applications, such as steam-generation power plants, petroleum, and other chemical plants. Also, by using the test fluid R-113, the Pool Boiling Experiment can provide some basic understanding of the boiling behavior of cryogenic fluids without the large cost of an experiment using an actual cryogen.

  19. Conceptual design for spacelab pool boiling experiment

    NASA Technical Reports Server (NTRS)

    Lienhard, J. H.; Peck, R. E.

    1978-01-01

    A pool boiling heat transfer experiment to be incorporated with a larger two-phase flow experiment on Spacelab was designed to confirm (or alter) the results of earth-normal gravity experiments which indicate that the hydrodynamic peak and minimum pool boiling heat fluxes vanish at very low gravity. Twelve small sealed test cells containing water, methanol or Freon 113 and cylindrical heaters of various sizes are to be built. Each cell will be subjected to one or more 45 sec tests in which the surface heat flux on the heaters is increased linearly until the surface temperature reaches a limiting value of 500 C. The entire boiling process will be photographed in slow-motion. Boiling curves will be constructed from thermocouple and electric input data, for comparison with the motion picture records. The conduct of the experiment will require no more than a few hours of operator time.

  20. SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

    DOEpatents

    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.

  1. Fundamental Boiling and RP-1 Freezing Experiments

    NASA Technical Reports Server (NTRS)

    Goode, Brian

    2002-01-01

    The prestart thermal conditioning of the hardware in LOX (liquid oxygen) systems involve heat transfer between LOX and metal where boiling plays a large role. Information is easily found on nucleate boiling, maximum heat flux, minimum heat flux and film boiling for common fluids like water. After looking at these standard correlations it was felt more data was needed for the cool down side transition boiling for the LN2 and LOX. In particular interest is the film boiling values, the temperature at which transition begins and the slope as peak heat flux is approached. The ultimate goal is an array of boiling heat transfer coefficient as a function of surface temperature which can be used in the chilldown model of the feed system, engine and bleed system for X-34. The first experiment consisted of an actual MC-1 LOX Impeller which had been machined backwards, that was instrumented with 17 surface thermocouples and submerged in liquid nitrogen. The thermocouples were installed on metal thicknesses varying from the thin inducer to the thick hub.

  2. SAFER03 and TRAC-BD1 analyses of a ROSA-III large-break experiment on a boiling water reactor

    SciTech Connect

    Itoya, S.; Kato, M.; Abe, N.; Nagasaka, I.

    1987-04-01

    Simulation tests of a boiling water reactor (BWR) loss-of-coolant accident (LOCA) caused by a double-ended guillotine break of a recirculation pump suction line were carried out with the rig of safety assessment III (ROSA-III) test facility at the Japan Atomic Energy Research Institute. SAFER03 and TRAC-BD1 analyses with the ROSA-III test data have been performed to assess predictive capability for the large-break LOCA thermal-hydraulic response of a BWR. The analytical results indicate that SAFER03 and TRAC-BD1 predicted key phenomena very well. Furthermore, it was confirmed that SAFER03 predicted higher peak cladding surface temperature than the test data due to the conservative thermal-hydraulic model.

  3. Fundamental Boiling and RP-1 Freezing Experiments

    NASA Technical Reports Server (NTRS)

    Goode, Brian; Turner, Larry D. (Technical Monitor)

    2001-01-01

    This paper describes results from experiments performed to help understand certain aspects of the MC-1 engine prestart thermal conditioning procedure. The procedure was constrained by the fact that the engine must chill long enough to get quality LOX at the LOX pump inlet but must be short enough to prevent freezing of RP-1 in the fuel pump. A chill test of an MC-1 LOX impeller was performed in LN2 to obtain data on film boiling, transition boiling and impeller temperature histories. The transition boiling data was important to the chill time so a subsequent experiment was performed chilling simple steel plates in LOX to obtain similar data for LOX. To address the fuel freezing concern, two experiments were performed. First, fuel was frozen in a tray and its physical characteristics were observed and temperatures of the fuel were measured. The result was physical characteristics as a function of temperature. Second was an attempt to measure the frozen thickness of RP-1 on a cold wall submerged in warm RP-1 and to develop a method for calculating that thickness for other conditions.

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

    DOEpatents

    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.

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

    DOEpatents

    Forsberg, Charles W.

    1987-01-01

    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 (non-borated) 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 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.

  6. Zero Boil-Off Tank (ZBOT) Experiment

    NASA Technical Reports Server (NTRS)

    Mcquillen, John

    2016-01-01

    The Zero-Boil-Off Tank (ZBOT) experiment has been developed as a small scale ISS experiment aimed at delineating important fluid flow, heat and mass transport, and phase change phenomena that affect cryogenic storage tank pressurization and pressure control in microgravity. The experiments use a simulant transparent low boiling point fluid (PnP) in a sealed transparent Dewar to study and quantify: (a) fluid flow and thermal stratification during pressurization; (b) mixing, thermal destratification, depressurization, and jet-ullage penetration during pressure control by jet mixing. The experiment will provide valuable microgravity empirical two-phase data associated with the above-mentioned physical phenomena through highly accurate local wall and fluid temperature and pressure measurements, full-field phase-distribution and flow visualization. Moreover, the experiments are performed under tightly controlled and definable heat transfer boundary conditions to provide reliable high-fidelity data and precise input as required for validation verification of state-of-the-art two-phase CFD models developed as part of this research and by other groups in the international scientific and cryogenic fluid management communities.

  7. GE simplified boiling water reactor stability analysis in time domain

    NASA Astrophysics Data System (ADS)

    Lu, Shanlai

    1997-12-01

    General Electric Simplified Boiling Water Reactor (SBWR) was designed as a next generation light water reactor. It uses natural circulation to remove the heat from the reactor core. Because of this unique in-vessel circulation feature, SBWR is expected to exhibit different stability behaviors. The main emphasis of this thesis is to study the SBWR stability behavior in the time domain. The best-estimate BWR accident/transient analysis computer code, TRAC-BF1, is employed to analyze the SBWR stability behavior. A detailed TRAC-BF1 SBWR model has been developed, which has the capability to model the in-vessel natural circulation and the reactor core kinetics. The model is used to simulate three slow depressurization processes. The simulation results show that the reactor is stable under low pressure and nominal downcomer water level conditions. However, when the downcomer water level is raised to about 19.2 m above the bottom of the reactor vessel, an unstable power oscillation is observed. The identified power oscillation is further analyzed using TRAC-BF1 1-D kinetics and the new TRAC-BF1 3-D kinetics code developed in this thesis. The effects of different time step sizes and vessel model nodalizations are examined. It is found that the power oscillation is in-phase and has a frequency of 0.3 HZ. In order to further explore the physical instabilty initiation mechanisms, a simplified dynamic model consisting of six simple differential equations is developed. The simplified model is able to predict the dominant physical phenomenon identified by the TRAC-BF1 analysis. The results indicate that the system instability is possibly caused by the steam separator hydro-static head oscillation under the high water level condition. In order to explore the higher order spacial effect of power oscillation, a 3-D reactor core kinetics code is coupled with the TRAC-BF1 computer code in the PVM parallel processing environment. A new coupling scheme and a multiple time step marching

  8. Aging study of boiling water reactor high pressure injection systems

    SciTech Connect

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  9. Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    SciTech Connect

    Chu, T.Y.; Slezak, S.E.; Bentz, J.H.; Pasedag, W.F.

    1994-03-01

    This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm{sup 2} across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests.

  10. Self-Sustaining Thorium Boiling Water Reactors

    SciTech Connect

    Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra; Seifried, Jeffrey E.; Zhang, Guanheng; Varela, Christopher R.; Fratoni, Massimiliano; Vijic, Jasmina J.; Downar, Thomas; Hall, Andrew; Ward, Andrew; Jarrett, Michael; Wysocki, Aaron; Xu, Yunlin; Kazimi, Mujid; Shirvan, Koroush; Mieloszyk, Alexander; Todosow, Michael; Brown, Nicolas; Cheng, Lap

    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.

  11. Peculiarities of the distribution of phases in the updraft section of a housed boiling reactor

    SciTech Connect

    Fedulin, V.N.; Bartolomei, G.G.; Shmelev, V.E.; Solodkii, V.A.

    1984-12-01

    This paper presents the results of an investigation of the structure of a two-phase flow in a large-diameter updraft section (D =2 m, H = 3 m) using the electroprobing method on a VK-50 boiling reactor. The structure of the reactor, its operating regimes, and the layout of the placement of local steam content sensors are described and the procedure for analysis of the experimental results is discussed.

  12. ASTRID: A 3D Eulerian software for subcooled boiling modelling - comparison with experimental results in tubes and annuli

    SciTech Connect

    Briere, E.; Larrauri, D.; Olive, J.

    1995-09-01

    For about four years, Electricite de France has been developing a 3-D computer code for the Eulerian simulation of two-phase flows. This code, named ASTRID, is based on the six-equation two-fluid model. Boiling water flows, such as those encountered in nuclear reactors, are among the main applications of ASTRID. In order to provide ASTRID with closure laws and boundary conditions suitable for boiling flows, a boiling model has been developed by EDF and the Institut de Mecanique des Fluides de Toulouse. In the fluid, the heat and mass transfer between a bubble and the liquid is being modelled. At the heating wall, the incipient boiling point is determined according to Hsu`s criterion and the boiling heat flux is split into three additive terms: a convective term, a quenching term and a vaporisation term. This model uses several correlations. EDF`s program in boiling two-phase flows also includes experimental studies, some of which are performed in collaboration with other laboratories. Refrigerant subcooled boiling both in tubular (DEBORA experiment, CEN Grenoble) and in annular geometry (Arizona State University Experiment) have been computed with ASTRID. The simulations show the satisfactory results already obtained on void fraction and liquid temperature. Ways of improvement of the model are drawn especially on the dynamical part.

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

    SciTech Connect

    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.

  14. Boils

    MedlinePlus

    ... the boil is very bad or comes back. Antibacterial soaps and creams cannot help much once a boil ... following may help prevent the spread of infection: Antibacterial soaps Antiseptic (germ-killing) washes Keeping clean (such as ...

  15. Boiling Experiment Facility for Heat Transfer Studies in Microgravity

    NASA Technical Reports Server (NTRS)

    Delombard, Richard; McQuillen, John; Chao, David

    2008-01-01

    Pool boiling in microgravity is an area of both scientific and practical interest. By conducting tests in microgravity, it is possible to assess the effect of buoyancy on the overall boiling process and assess the relative magnitude of effects with regards to other "forces" and phenomena such as Marangoni forces, liquid momentum forces, and microlayer evaporation. The Boiling eXperiment Facility is now being built for the Microgravity Science Glovebox that will use normal perfluorohexane as a test fluid to extend the range of test conditions to include longer test durations and less liquid subcooling. Two experiments, the Microheater Array Boiling Experiment and the Nucleate Pool Boiling eXperiment will use the Boiling eXperiment Facility. The objectives of these studies are to determine the differences in local boiling heat transfer mechanisms in microgravity and normal gravity from nucleate boiling, through critical heat flux and into the transition boiling regime and to examine the bubble nucleation, growth, departure and coalescence processes. Custom-designed heaters will be utilized to achieve these objectives.

  16. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    SciTech Connect

    Boing, L.E.

    1989-12-01

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs.

  17. Feasibility study on the thorium fueled boiling water breeder reactor

    SciTech Connect

    PetrusTakaki, N.

    2012-07-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  18. Generic safety insights for inspection of boiling water reactors

    SciTech Connect

    Higgins, J.C.; Taylor, J.H.; Fresco, A.N.; Hillman, B.M.

    1987-01-01

    As the number of operating nuclear power plants (NPPs) increases, safety inspection has increased in importance. Over the last 2 yr, probabilistic risk assessment (PRA) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only approx. 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the Brookhaven National Lab. program to develop generic boiling water reactor (BWR) PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs.

  19. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    DOEpatents

    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)

  20. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    DOEpatents

    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)

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

    NASA Astrophysics Data System (ADS)

    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.

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

    DOE PAGESBeta

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

    2016-06-04

    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 themore » 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.« less

  3. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-31

    ... 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... modifying or adding EP requirements in Section 50.47, Section 50.54, and Appendix E of 10 CFR part 50 (76...

  4. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-08

    ... revised 10 CFR 73.55 through the issuance of a final rule on March 27, 2009 (74 FR 13926). Section 73.55... 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...

  5. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-03

    ... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas and... Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Flint North building, 11555 Rockville Pike, Rockville, MD. After registering with security,...

  6. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-27

    ..., which utilized a forced-circulation, direct-cycle boiling water reactor as its heat source. The plant is... March 27, 2009 (74 FR 13926). The revised regulation stated that it was applicable to all Part 50... COMMISSION Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain...

  7. Observations of the boiling process from a downward-facing torispherical surface: Confirmatory testing of the heavy water new production reactor flooded cavity design

    SciTech Connect

    Chu, T.Y.; Bentz, J.H.; Simpson, R.B.

    1995-06-01

    Reactor-scale ex-vessel boiling experiments were performed in the CYBL facility at Sandia National Laboratories. The boiling flow pattern outside the RPV bottom head shows a center pulsating region and an outer steady two-phase boundary layer region. The local heat transfer data can be correlated in terms of a modified Rohsenow correlation.

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

    DOEpatents

    Gluntz, D.M.

    1996-03-12

    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.

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

    DOEpatents

    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.

  10. Camera Inspection Arm for Boiling Water Reactors - 13330

    SciTech Connect

    Martin, Scott; Rood, Marc

    2013-07-01

    Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

  11. CFD simulation of DEBORA boiling experiments

    NASA Astrophysics Data System (ADS)

    Rzehak, Roland; Krepper, Eckhard

    2012-08-01

    In this work we investigate the present capabilities of computational fluid dynamics for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. This kind of modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant non-dimensional numbers have been realized in the DEBORA tests using dichlorodifluoromethane (R12) as the working fluid. This facilitated measurements of radial profiles for gas volume fraction, gas velocity, bubble size and liquid temperature as well as axial profiles of wall temperature. After reviewing the theoretical and experimental basis of correlations used in the ANSYS CFX model used for the calculations, we give a careful assessment of the necessary recalibrations to describe the DEBORA tests. The basic CFX model is validated by a detailed comparison to the experimental data for two selected test cases. Simulations with a single set of calibrated parameters are found to give reasonable quantitative agreement with the data for several tests within a certain range of conditions and reproduce the observed tendencies correctly. Several model refinements are then presented each of which is designed to improve one of the remaining deviations between simulation and measurements. Specifically we consider a homogeneous MUSIG model for the bubble size, modified bubble forces, a wall function for turbulent boiling flow and a partial slip boundary condition for the liquid phase. Finally, needs for further model developments are identified and promising directions discussed.

  12. Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

    SciTech Connect

    Yin, Shengjun; Bass, Bennett Richard; Stevens, Gary; Kirk, Mark

    2011-01-01

    This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle3 3. PWR inlet nozzle; ; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; To assess the significance of attached piping loads on the stresses in the nozzle corner region; and To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

  13. Boiling-Water Reactor internals aging degradation study. Phase 1

    SciTech Connect

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

  14. Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

    SciTech Connect

    Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

    1990-11-01

    The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs.

  15. Models and Stability Analysis of Boiling Water Reactors

    SciTech Connect

    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

  16. Industrial application of APOLLO2 to boiling water reactors

    SciTech Connect

    Marotte, V.; Clement, F.; Thareau, S.; Misu, S.; Zmijarevic, I.

    2006-07-01

    AREVA NP - a joint's subsidiary of AREVA and Siemens- decided to develop a new calculation scheme based on the multigroup neutron transport code APOLLO2, developed at CEA, for industrial application to Boiling Water Reactors. This scheme is based on the CEA93 library with the XMAS-172 energy mesh and the JEF2.2 evaluation. Microscopic cross-sections are improved by a self-shielding calculation that accounts for 2D geometrical effects and the overlapping of resonances. The flux is calculated with the Method of Characteristics. A best-estimate flux is found with the 172 energy group structure. In the industrial scheme, the computing time and the memory size are reduced by a simplified self-shielding and the calculation of the flux with 26 energy groups. The results are presented for three BWR assemblies. Several BWR operating conditions were simulated. Results are accurate compared to the Monte-Carlo code MCNP. A very good agreement is obtained between the best-estimate and the industrial calculations, also during depletion. These results show the high physical quality of the APOLLO2 code and its capability to calculate accurately BWR assemblies for industrial applications. (authors)

  17. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  18. Improvements of fuel failure detection in boiling water reactors using helium measurements

    SciTech Connect

    Larsson, I.; Sihver, L.; Grundin, A.; Helmersson, J. O.

    2012-07-01

    To certify a continuous and safe operation of a boiling water reactor, careful surveillance of fuel integrity is of high importance. The detection of fuel failures can be performed by off-line gamma spectroscopy of off-gas samples and/or by on-line nuclide specific monitoring of gamma emitting noble gases. To establish the location of a leaking fuel rod, power suppression testing can be used. The accuracy of power suppression testing is dependent on the information of the delay time and the spreading of the released fission gases through the systems before reaching the sampling point. This paper presents a method to improve the accuracy of power suppression testing by determining the delay time and gas spreading profile. To estimate the delay time and examine the spreading of the gas in case of a fuel failure, helium was injected in the feed water system at Forsmark 3 nuclear power plant. The measurements were performed by using a helium detector system based on a mass spectrometer installed in the off-gas system. The helium detection system and the results of the experiment are presented in this paper. (authors)

  19. Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor

    NASA Astrophysics Data System (ADS)

    Cheng, Ling

    The thesis evaluates the safety systems in the next generation boiling water reactor by analyzing the main steam line break loss of coolant accident performed in the Purdue university multi-dimensional test assembly (PUMA). RELAP5 code simulations, both for the PUMA main steam line break (MSLB) case and for the simplified boiling water reactor (SBWR) MSLB case have been utilized to compare with the experiment data. The comparison shows that RELAP5 is capable to perform the safety analysis for SBWR. The comparison also validates the three-level scaling methodology applied to the design of the PUMA facility. The PUMA suppression pool mixing and condensation test data have been studied to give the detailed understanding on this important local phenomenon. A simple one dimensional integral model, which can reasonably simulate the mixing process inside suppression pool have been developed and the comparison between the model prediction and the experiment data demonstrates the model can be utilized for analyzing the suppression pool mixing process.

  20. (Installation of a boiling water reactor core melt progression phenomena program)

    SciTech Connect

    Ott, L.J.

    1990-06-07

    The CORA operational staff at Kernforschungszentrum Karlsruhe (KfK) requested, under the auspices of the Severe Fuel Damage Partners Program, that Oak Ridge National Laboratory (ORNL) developed models, specific to boiling water reactor (BWR) response under severe accident conditions, be applied in support of future BWR experiments to be performed in the CORA facility. Accordingly, the current Statement of Work for the BWR Core Melt Progression Phenomena Program provides for the development of a CORA-specific BWR experimental model to analyze the results of CORA BWR experiments and the planning of future experiments. The traveler installed version 1.0 of the CORA/BWR experiment-specific code on KfK personal computers and assisted the CORA staff in their preliminary pretest analyses for CORA test 18.

  1. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    SciTech Connect

    Arcieri, W.C.; Hanson, D.J. )

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents.

  2. Aging assessment and license renewals: Plant life management for the first stage boiling water reactor

    SciTech Connect

    Tezuka, Kenichi; Kawamura, Shinichi; Aoki, Masataka; Mori, Tsuguo

    1996-09-01

    The first stage Boiling Water Reactors (BWRs) have been operating more than 25 years. Some components have potential of failure by aging. So, evaluations have been done for the main components such as Reactor Pressure Vessel (RPV), Reactor Internals, Primary Coolant Piping, Reactor Recirculation Pump, Cable (Inside PCV), Primary Containment Vessel (PCV) and Concrete Structure. This evaluation has been done by joint study between electric utilities and manufacturers to confirm integrity and identify necessary development.

  3. Reactor antineutrino experiments

    NASA Astrophysics Data System (ADS)

    Lu, Haoqi

    2014-05-01

    Neutrinos are elementary particles in the standard model of particle physics. There are three flavors of neutrinos that oscillate among themselves. Their oscillation can be described by a 3×3 unitary matrix, containing three mixing angles θ12, θ23, θ13, and one CP phase. Both θ12 and θ23 are known from previous experiments. θ13 was unknown just two years ago. The Daya Bay experiment gave the first definitive non-zero value in 2012. An improved measurement of the oscillation amplitude sin 22(θ 13) = 0.090+0.008-0.009 and the first direct measurement of the \\bar ν e mass-squared difference ∣ Δ m2ee∣ = (2.59+0.19-0.20)× 10-3 eV2 were obtained recently. The large value of θ13 boosts the next generation of reactor antineutrino experiments designed to determine the neutrino mass hierarchy, such as JUNO and RENO-50.

  4. HEDL W-1 SLSF experiment LOPI transient and boiling test results. [LMFBR

    SciTech Connect

    Henderson, J.M.; Wood, S.A.; Rothrock, R.B.

    1980-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment was designed to study the heat release characteristics of fast reactor fuel pins under Loss-of-Piping-Integrity (LOPI) accident conditions and determine stable sodium boiling initiation and recovery limits in a prototypic fuel pin bundle array. The results of the experiment address major second level of assurance (LOA-2) safety issues and provide increased insight and understanding of phenomena that would inherently terminate hypothesized accidents with only limited core damage. The irradiation phase of the experiment, consisting of thirteen individual transients, was performed between May 27 and July 20, 1979. The final transient produced approximately two seconds of coolant boiling, cladding dryout, and incipient fuel pin failure. The facility and test hardware performed as designed, allowing completion of all planned tests and achievement of all test objectives.

  5. ATWS Analysis with an Advanced Boiling Curve Approach within COBRA 3-CP

    SciTech Connect

    Gensler, A.; Knoll, A.; Kuehnel, K.

    2007-07-01

    In 2005 the German Reactor Safety Commission issued specific requirements on core coolability demonstration for PWR ATWS (anticipated transients without scram). Thereupon AREVA NP performed detailed analyses for all German PWRs. For a German KONVOI plant the results of an ATWS licensing analysis are presented. The plant dynamic behavior is calculated with NLOOP, while the hot channel analysis is performed with the thermal hydraulic computer code COBRA 3-CP. The application of the fuel rod model included in COBRA 3-CP is essential for this type of analysis. Since DNB (departure from nucleate boiling) occurs, the advanced post DNB model (advanced boiling curve approach) of COBRA 3-CP is used. The results are compared with those gained with the standard BEEST model. The analyzed ATWS case is the emergency power case 'loss of main heat sink with station service power supply unavailable'. Due to the decreasing coolant flow rate during the transient the core attains film boiling conditions. The results of the hot channel analysis strongly depend on the performance of the boiling curve model. The BEEST model is based on pool boiling conditions whereas typical PWR conditions - even in most transients - are characterized by forced flow for which the advanced boiling curve approach is particularly suitable. Compared with the BEEST model the advanced boiling curve approach in COBRA 3-CP yields earlier rewetting, i.e. a shorter period in film boiling. Consequently, the fuel rod cladding temperatures, that increase significantly due to film boiling, drop back earlier and the high temperature oxidation is significantly diminished. The Baker-Just-Correlation was used to calculate the value of equivalent cladding reacted (ECR), i.e. the reduction of cladding thickness due to corrosion throughout the transient. Based on the BEEST model the ECR value amounts to 0.4% whereas the advanced boiling curve only leads to an ECR value of 0.2%. Both values provide large margins to the 17

  6. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-22

    ... October 14, 2009 (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available on... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR) The ACRS Subcommittee on ABWR will hold a meeting on March 2, 2010, at...

  7. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-09

    ... October 14, 2009, (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available on... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will hold a meeting on March...

  8. Natural Convection and Boiling for Cooling SRP Reactors During Loss of Circulation Conditions

    SciTech Connect

    Buckner, M.R.

    2001-06-26

    This study investigated natural convection and boiling as a means of cooling SRP reactors in the event of a loss of circulation accident. These studies show that single phase natural convection cooling of SRP reactors in shutdown conditions with the present piping geometry is probably not feasible.

  9. Future reactor experiments

    NASA Astrophysics Data System (ADS)

    Wen, Liangjian

    2015-07-01

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  10. Future reactor experiments

    SciTech Connect

    Wen, Liangjian

    2015-07-15

    The non-zero neutrino mixing angle θ{sub 13} has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  11. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    SciTech Connect

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  12. A model for stationary and dynamic impression of undercooled boiling in coupled thermohydraulic and neutron physics calculations of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Mueller, Roland Guenther

    1987-06-01

    In order to account for subcooled boiling in calculations of neutron physics and thermal hydraulics of light water reactors (where vapor bubbles strongly influence the nuclear chain reaction), a dynamic model is derived from the time-dependent conservation equations. It contains methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. It enables the complete two-phase flow region to be treated consistently. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement is reached. The results from the coupling of the new calculation model with a neutron kinetics program proves its suitability for the steady-state and transient calculation of reactor cores.

  13. Packed Bed Reactor Experiment

    NASA Video Gallery

    The purpose of the Packed Bed Reactor Experiment in low gravity is to determine how a mixture of gas and liquid flows through a packed bed in reduced gravity. A packed bed consists of a metal pipe ...

  14. Aging assessment of the boiling-water reactor (BWR) standby liquid control system

    SciTech Connect

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.

  15. Aging assessment of the boiling-water reactor (BWR) standby liquid control system. Phase 1

    SciTech Connect

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.

  16. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect

    Ishii, M.; Xu, Y.; Revankar, S.T.

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  17. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Classification of decommissioning wastes. Addendum 2

    SciTech Connect

    Murphy, E.S.

    1984-09-01

    The radioactive wastes expected to result from decommissioning of the reference boiling water reactor power station are reviewed and classified in accordance with 10 CFR 61. The 18,949 cubic meters of waste from DECON are classified as follows: Class A, 97.5%; Class B, 2.0%; Class C, 0.3%. About 0.2% (47 cubic meters) of the waste would be generally unacceptable for disposal using near-surface disposal methods.

  18. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    SciTech Connect

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  19. Electrical design of payload G-534: The Pool Boiling Experiment

    NASA Technical Reports Server (NTRS)

    Francisco, David R.

    1992-01-01

    Payload G-534, the Pool Boiling Experiment (PBE), is a Get Away Special that is scheduled to fly on the shuttle in 1992. This paper will give a brief overall description of the experiment with the main discussion being the electrical design with a detailed description of the power system and interface to the GAS electronics. The batteries used and their interface to the experiment Power Control Unit (PCU) and GAS electronics will be examined. The design philosophy for the PCU will be discussed in detail. The criteria for selection of fuses, relays, power semiconductors and other electrical components along with grounding and shielding policy for the entire experiment will be presented. The intent of this paper is to discuss the use of military tested parts and basic design guidelines to build a quality experiment for minimal additional cost.

  20. Reactor antineutrino experiments

    NASA Astrophysics Data System (ADS)

    Lu, Haoqi

    2014-09-01

    Neutrinos are elementary particles in the standard model of particle physics. There are three flavors of neutrinos that oscillate among themselves. Their oscillation can be described by a 3×3 unitary matrix, containing three mixing angles θ12, θ23, θ13, and one CP phase. Both θ12 and θ23 are known from previous experiments. θ13 was unknown just two years ago. The Daya Bay experiment gave the first definitive nonzero value in 2012. An improved measurement of the oscillation amplitude sin 22(θ 13) = 0.090+0.008-0.009 and the first direct measurement of the \\bar ν e mass-squared difference \\vertΔ m2ee\\vert\\big (2.59+0.19-0.20\\big )×10-3 eV2 were obtained recently. The large value of θ13 boosts the next generation of reactor antineutrino experiments designed to determine the neutrino mass hierarchy, such as JUNO and RENO-50.

  1. Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

    SciTech Connect

    Chandler, David; Primm, Trent; Maldonado, G Ivan

    2010-01-01

    The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction offlux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. 'Foils' (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil. This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil

  2. A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

    NASA Technical Reports Server (NTRS)

    Sorensen, Kirk; Juhasz, Albert

    2007-01-01

    Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

  3. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    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 Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for...

  4. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-27

    ..., which utilized a forced-circulation, direct-cycle boiling water reactor as its heat source. The plant is... March 27, 2009 (74 FR 13926). The revised regulation stated that it was applicable to all Part 50... published in the Federal Register on May 8, 2012 (77 FR 27097). Based upon the environmental assessment,...

  5. Development and Capabilities of ISS Flow Boiling and Condensation Experiment

    NASA Technical Reports Server (NTRS)

    Nahra, Henry; Hasan, Mohammad; Balasubramaniam, R.; Patania, Michelle; Hall, Nancy; Wagner, James; Mackey, Jeffrey; Frankenfield, Bruce; Hauser, Daniel; Harpster, George; Nawrocki, David; Clapper, Randy; Kolacz, John; Butcher, Robert; May, Rochelle; Chao, David; Mudawar, Issam; Kharangate, Chirag R.; O'Neill, Lucas E.

    2015-01-01

    An experimental facility to perform flow boiling and condensation experiments in long duration microgravity environment is being designed for operation on the International Space Station (ISS). This work describes the design of the subsystems of the FBCE including the Fluid subsystem modules, data acquisition, controls, and diagnostics. Subsystems and components are designed within the constraints of the ISS Fluid Integrated Rack in terms of power availability, cooling capability, mass and volume, and most importantly the safety requirements. In this work we present the results of ground-based performance testing of the FBCE subsystem modules and test module which consist of the two condensation modules and the flow boiling module. During this testing, we evaluated the pressure drop profile across different components of the fluid subsystem, heater performance, on-orbit degassing subsystem, heat loss from different modules and components, and performance of the test modules. These results will be used in the refinement of the flight system design and build-up of the FBCE which is manifested for flight in late 2017-early 2018.

  6. 3. RW Meyer Sugar Mill: 18761889. Sorghum pan and boiling ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    3. RW Meyer Sugar Mill: 1876-1889. Sorghum pan and boiling range flue. Manufactured by John Nott & Co., Honolulu, Hawaii, 1878. View: South side of sorghum pan and boiling range flue. In the sorghum pan heat was applied to the cane juice to clarify it, evaporate its water content, and concentrate the sugar crystals. Hot gasses moved through the flue underneath the entire copper bottom of the sorghum pan from the furnace (east) end to the smokestack (west) end of the boiling range. The sorghum pan sides are of redwood. The flue is built of fire-brick, masonry, and portland cement. - R. W. Meyer Sugar Mill, State Route 47, Kualapuu, Maui County, HI

  7. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    SciTech Connect

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  8. Generic risk insights for General Electric boiling water reactors

    SciTech Connect

    Travis, R.; Taylor, J. ); Chung, J. . Office of Nuclear Reactor Regulation)

    1991-05-01

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of General Electric boiling water rectors and applying the insights gained to plants that have not been subjected to a PRA. The available risk assessments (six plants) were examined to identify the most probable, i.e., dominant accident sequences at each plants. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eight sequences met this definition. From these sequences, the most important component failures and human error that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training, maintenance, design review, and inspections. 13 refs., 6 tabs.

  9. Pulsed reactor experiments at Oak Ridge

    SciTech Connect

    Mihalczo, J.T.

    1994-12-31

    This paper describes dynamic experiments for 3 pulsed reactors. 1st reactor was pulsed from some average power by rotating a partial Be reflector past an unreflected core face; the other 2 reactors were pulsed by rapid insertion of a fuel rod into the unmoderated and unreflected reactor at essentially zero neutron level with no significant neutron source present. The first reactor was a mockup of an EURATOM design (never constructed) of the proposed SORA Reactor, and the other two were the Health Physics Research Reactor and the Army Pulse Radiation Facility Reactor (APRFR). This paper describes the experiments performed in initial testing of these systems, including destructive tests of APRFR, to set operating limits for this type of reactor in pulsed operation. All the experiments described were performed at the Oak Ridge Critical Experiments Facility.

  10. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Design Certification Rule for the U.S. Advanced Boiling Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS Pt. 52, App. A Appendix A to Part 52—Design Certification Rule for the U.S. Advanced Boiling...

  11. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Design Certification Rule for the U.S. Advanced Boiling Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS Pt. 52, App. A Appendix A to Part 52—Design Certification Rule for the U.S. Advanced Boiling...

  12. Conceptual design and thermal-hydraulic characteristics of natural circulation Boiling Water Reactors

    SciTech Connect

    Kataoka, Y.; Suzuki, H.; Murase, M. ); Horiuchi, T.; Miki, M. )

    1988-08-01

    A natural circulation boiling water reactor (BWR) with a rated capacity of 600 MW (electric) has been conceptually designed for small- and medium-sized light water reactors. The components and systems in the reactor are simplified by eliminating pumped recirculation systems and pumped emergency core cooling systems. Consequently, the volume of the reactor building is -- 50% of that for current BWRs with the same rated capacity; the construction period is also shorter. Its thermal-hydraulic characteristics, critical power ratio (CPR) and flow stability at steady state, decrease in the minimum CPR (..delta..MCPR) at transients, and the two-phase mixture level in the reactor pressure vessel (RPV) during accidents are investigated. The two-phase mixture level in the RPV during an accident does not decrease to lower than the top of the core; the core uncovery and heatup of fuel cladding would not occur during any loss-of-coolant accident.

  13. Boiling Water Reactor Fuel Cycle Optimization for Prevention of Channel-Blade Interference

    SciTech Connect

    Kropaczek, David J.; Karve, Atul A.; Oyarzun, Christian C.; Asgari, Mehdi

    2006-07-01

    A formal optimization method for eliminating the potential of Boiling Water Reactor channel-blade interference is presented within the context of fuel cycle design. The method is based on the use of threshold constraints on blade force as penalty terms within an objective function that are employed as part of a search algorithm. Results demonstrate the effectiveness of the constraint formulation in eliminating channel-blade interference as part of the design of the core loading and operational strategy. (authors)

  14. Optimal control of a boiling water reactor load-following operation

    SciTech Connect

    Lin, C.; Lin, Z.P.; Jiang, W.J. . Dept. of Nuclear Engineering)

    1989-06-01

    The authors describe a method based on a forward dynamic programming technique applied to load-following control of a boiling water reactor. The control strategy obtained is optimal and satisfies operation constraints. A course-mesh, one-dimensional model using the two-group diffusion theory with Doppler, void, and xenon feedbacks is developed to reduce computer time. The control rods are assumed to be fixed during load maneuvers, and variations in core power are accomplished through core flow.

  15. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-16

    ...The U.S. Nuclear Regulatory Commission (NRC or the Commission) is amending its regulations to certify an amendment to the U.S. Advanced Boiling Water Reactor (U.S. ABWR) standard plant design to comply with the NRC's aircraft impact assessment (AIA) regulations. This action allows applicants or licensees intending to construct and operate a U.S. ABWR to comply with the NRC's AIA regulations by......

  16. Severe accident sequence assessment for boiling water reactors: program overview

    SciTech Connect

    Fontana, M. H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case.

  17. Towards testing the unitarity of the 3 × 3 lepton flavor mixing matrix in a precision reactor antineutrino oscillation experiment

    NASA Astrophysics Data System (ADS)

    Xing, Zhi-zhong

    2013-01-01

    The 3 × 3 Maki-Nakagawa-Sakata-Pontecorvo (MNSP) lepton flavor mixing matrix may be slightly non-unitary if the three active neutrinos are coupled with sterile neutrinos. We show that it is in principle possible to test whether the relation |Ve1 | 2 +|Ve2 | 2 +|Ve3 | 2 = 1 holds or not in a precision reactor antineutrino oscillation experiment, such as the recently proposed Daya Bay II experiment. We explore three categories of non-unitary effects on the 3 × 3 MNSP matrix: 1) the indirect effect in the (3 + 3) flavor mixing scenario where the three heavy sterile neutrinos do not take part in neutrino oscillations; 2) the direct effect in the (3 + 1) scenario where the light sterile neutrino can oscillate into the active ones; and 3) the interplay of both of them in the (3 + 1 + 2) scenario. We find that both the zero-distance effect and flavor mixing factors of different oscillation modes can be used to determine or constrain the sum of |Ve1 | 2, |Ve2 | 2 and |Ve3 | 2 and its possible deviation from one, and the active neutrino mixing angles θ12 and θ13 can be cleanly extracted even in the presence of light or heavy sterile neutrinos. Some useful analytical results are obtained for each of the three scenarios.

  18. Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

    SciTech Connect

    Revankar, S.T.; Xu, Y.; Yoon, H.J.; Ishii, M.

    2002-07-01

    The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance. (authors)

  19. Development of Flow Boiling and Condensation Experiment on the International Space Station- Normal and Low Gravity Flow Boiling Experiment Development and Test Results

    NASA Technical Reports Server (NTRS)

    Nahra, Henry K.; Hall, Nancy R.; Hasan, Mohammad M.; Wagner, James D.; May, Rochelle L.; Mackey, Jeffrey R.; Kolacz, John S.; Butcher, Robert L.; Frankenfield, Bruce J.; Mudawar, Issam; Konichi, Chris; Hyounsoon, Lee

    2013-01-01

    Flow boiling and condensation have been identified as two key mechanisms for heat transport that are vital for achieving weight and volume reduction as well as performance enhancement in future space systems. Since inertia driven flows are demanding on power usage, lower flows are desirable. However, in microgravity, lower flows are dominated by forces other than inertia (like the capillary force). It is of paramount interest to investigate limits of low flows beyond which the flow is inertial enough to be gravity independent. One of the objectives of the Flow Boiling and Condensation Flight Experiment sets to investigate these limits for flow boiling and condensation. A two-phase flow loop consisting of a Flow Boiling Module and two Condensation Modules has been developed to experimentally study flow boiling condensation heat transfer in the reduced gravity environment provided by the reduced gravity platform. This effort supports the development of a flow boiling and condensation facility for the International Space Station (ISS). The closed loop test facility is designed to deliver the test fluid, FC-72 to the inlet of any one of the test modules at specified thermodynamic and flow conditions. The zero-g-aircraft tests will provide subcooled and saturated flow boiling critical heat flux and flow condensation heat transfer data over wide range of flow velocities. Additionally, these tests will verify the performance of all gravity sensitive components, such as evaporator, condenser and accumulator associated with the two-phase flow loop. We will present in this paper the breadboard development and testing results which consist of detailed performance evaluation of the heater and condenser combination in reduced and normal gravity. We will also present the design of the reduced gravity aircraft rack and the results of the ground flow boiling heat transfer testing performed with the Flow Boiling Module that is designed to investigate flow boiling heat transfer and

  20. Environmental qualification testing of the prototype pool boiling experiment

    NASA Technical Reports Server (NTRS)

    Sexton, J. Andrew

    1992-01-01

    The prototype Pool Boiling Experiment (PBE) flew on the STS-47 mission in September 1992. This report describes the purpose of the experiment and the environmental qualification testing program that was used to prove the integrity of the prototype hardware. Component and box level vibration and thermal cycling tests were performed to give an early level of confidence in the hardware designs. At the system level, vibration, thermal extreme soaks, and thermal vacuum cycling tests were performed to qualify the complete design for the expected shuttle environment. The system level vibration testing included three axis sine sweeps and random inputs. The system level hot and cold soak tests demonstrated the hardware's capability to operate over a wide range of temperatures and gave the project team a wider latitude in determining which shuttle thermal altitudes were compatible with the experiment. The system level thermal vacuum cycling tests demonstrated the hardware's capability to operate in a convection free environment. A unique environmental chamber was designed and fabricated by the PBE team and allowed most of the environmental testing to be performed within the project's laboratory. The completion of the test program gave the project team high confidence in the hardware's ability to function as designed during flight.

  1. Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

    SciTech Connect

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1993-08-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set.

  2. Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

    SciTech Connect

    1994-04-30

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

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

    SciTech Connect

    Boing, L.E.; Henley, D.R. ); Manion, W.J.; Gordon, J.W. )

    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.

  4. OECD MCCI project enhancing instrumentation for reactor materials experiments, Rev. 0 September 3, 2002.

    SciTech Connect

    Lomperski, S.; Basu, S.

    2011-05-23

    Reactor safety experiments for studying the reactions of a molten core (corium) with water and/or concrete involve materials at extremely high temperature. Such high temperature severely restricts the types of sensors that can be employed to measure characteristics of the corium itself. Yet there is great interest in improving instrumentation so that the state of the melt can be established with more precision. In particular, it would be beneficial to increase both the upper range limit and accuracy of temperature measurements. The poor durability of thermocouples at high temperature is also an important issue. For experiments involving a water-quenched melt, direct measurements of the growth rate of the crust separating the melt and water would be of great interest. This is a key element in determining the nature of heat transfer between the melt and coolant. Despite its importance, no one has been able to directly measure the crust thickness during such tests. This paper considers three specialized sensors that could be introduced to enhance melt characterization: (1) A commercially fabricated, single point infrared temperature measurement with the footprint of a thermowell. A lens assembly and fiber optic cable linked to a receiver and amplifier measures the temperature at the base of a tungsten thermowell. The upper range limit is 3000 C and accuracy is {+-}0.25% of the reading. (2) In-house development of an ultrasonic temperature sensor that would provide multipoint measurements at temperatures up to {approx}3000 C. The sensors are constructed from tungsten rods and have a high temperature durability that is superior to that of thermocouples. (3) In-house development of an ultrasonic probe to measure the growth rate of the corium crust. This ultrasonic sensor would include a tungsten waveguide that transmits ultrasonic pulses up through the corium melt towards the crust and detects reflections from the melt/crust interface. A measurement of the echo time

  5. Boiling in variable gravity under the action of an electric field: results of parabolic flight experiments

    NASA Astrophysics Data System (ADS)

    Di Marco, P.; Raj, R.; Kim, J.

    2011-12-01

    Results from the variable gravity pool boiling experiments performed during the 52nd ESA parabolic flight campaign are reported in this paper. During a typical parabola, the gravity acceleration changes from 1.8gE (high gravity) to ~0gE (low gravity) and finally back to 1.8gE. The two high gravity periods and the microgravity period are each roughly maintained for 20 seconds while the transition from high gravity to low gravity and vice versa occurs over a period of 3-5 seconds. Use of the high feedback frequency microheater array allowed quasi-steady boiling data over the continuous range of gravity levels (0gE-1.8gE). The experimental apparatus consisted of a boiling chamber with a 7×7 mm2 microheater array in a 10×10 configuration. Each heater in the array was individually controlled to maintain a constant temperature. The array could be operated in a full configuration or a selectively powered reduced set of 3×3 heaters. Experiments were performed with FC-72 as the test fluid, the pressure was maintained at a constant value between 1 and 1.13 atm and the subcooling ranged from 27 to 11 K. An external electric field was imposed over the boiling surface by means of a grid consisting of 4 rods, laid parallel to the surface; voltages up to 10 kV were applied. The electric field was effective in reducing the size of the detaching bubbles, and increasing the heat transfer compared to the values in low-g, although its effectiveness decayed as the heat flux/superheat increased. The current results compared well with previous results obtained in the ARIEL apparatus that was operated in orbital flight.

  6. Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

    SciTech Connect

    Fellhauer, C.R.; Boing, L.E.; Aldana, J.

    1997-03-01

    The Final Report for the Decontamination and Decommissioning (D&D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D&D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D&D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D&D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a {open_quotes}Radiologically Controlled Area,{close_quotes} noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion).

  7. The Boiling eXperiment Facility (BXF) for the Microgravity Science Glovebox (MSG)

    NASA Technical Reports Server (NTRS)

    McQuillen, John; Chao, David; Vergilii, Frank

    2006-01-01

    Boiling is an effective means of cooling by removing heat from surfaces through vaporization of a working fluid. It is also affected by both the magnitude and direction of gravity. By conducting pool boiling tests in microgravity, the effect of buoyancy n the overall boiling process and the relative magnitude of other phenomena can be assessed. The Boiling eXperiment Facility (BXF) is being built for the Microgravity Science Glovebox. This facility will conduct two pool boiling studies. The first study the Microheater Array Boiling Experiment (MABE) uses two 96 element microheater arrays, 2.7 mm and 7.0 mm in size, to measure localized hear fluxes while operating at a constant temperature. The other experiment, the Nucleate Pool Boiling eXperiment (NPBX) uses a 85 mm diameter heater wafer that has been "seeded" with five individually-controlled nucleation sites to study bubble nucleation, growth, coalescence and departure. The BXF uses normal-perfluorohexane as the test fluid and will operate between pressures of 60 to 244 Pa. and temperatures of 35 to 60 C. Both sets of experimental heaters are highly instrumented. Pressure and bulk fluid temperature measurements will be made with standard rate video. A high speed video system will be used to visualize the boiling process through the bottom of the MABE heater arrays. The BXF is currently scheduled to fly on Utilization Flight-13A.1 to the ISS with facility integration into the MSG and operation during Increment 15

  8. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

    SciTech Connect

    Rosa, M.P.; Podowski, M.Z.

    1995-09-01

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are two-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.

  9. System for monitoring of energy release in the core of a boiling-water reactor

    SciTech Connect

    Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskii, I.I.

    1988-06-01

    Results are discussed from an investigation into a system for the physical monitoring of energy release in the core of the VK-50 boiling water reactor. Movable self-powered detectors are used in this system as energy-release neutron detectors. Rhodium serves as the emitter in these detectors. A number of parameters for these detectors were experimentally measured; they include the ratio of the instantaneous current to the activation current, the coefficients of relative sensitivity of the detectors to neutrons, and the effective cross sections for the interaction of /sup 103/Rh with thermal and epithermal neutrons. Computer simulation of the system and detectors is discussed. The system can monitor absolute energy distribution in the fuel assemblies under reactor conditions. The system is independent of thermotechnical measurements with an instrument to determine energy release in the reactor core.

  10. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    NASA Astrophysics Data System (ADS)

    Trianti, Nuri; Nurjanah, Su'ud, Zaki; Arif, Idam; Permana, Sidik

    2015-09-01

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid's temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

  11. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    SciTech Connect

    Trianti, Nuri Nurjanah,; Su’ud, Zaki; Arif, Idam; Permana, Sidik

    2015-09-30

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

  12. Pressure suppression containment system for boiling water reactor

    SciTech Connect

    Gluntz, D.M.; Nesbitt, L.B.

    1997-01-21

    A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

  13. ABWR (advanced boiling water reactor) Design Verification Program

    SciTech Connect

    Fox, J.N.

    1990-10-01

    The ABWR Design Verification Program is aimed at restoring confidence in the US licensing process by demonstrating its workability by obtaining USNRC preapproval of GE's ABWR Standard Plant. The purpose of this work is to achieve full NRC approval of the ABWR through the award of an NRC Staff final design approval (FDA) and design certification. The approach is to (1) establish a licensing basis with the NRC Staff for the ABWR, (2) prepare and submit, for NRC Staff review, an SSAR to obtain an FDA, and (3) participate in a rulemaking process to obtain certification of the ABWR design. This program was initiated August 27, 1986. This report, the fourth annual progress report, summarizes progress on this program from October 1, 1989 through September 30, 1990. 9 refs., 5 tabs.

  14. Risk based ISI application to a boiling water reactor

    SciTech Connect

    Smith, A.; Dimitrijevic, V.B.; O`Regan, P.J.

    1996-12-01

    The ASME Section XI Working Group on Implementation of Risk-Based Examination produced a code case to define risk-based selection rules that could be used for In-Service Inspection (ISI) of Class 1, 2, and 3 piping. To provide guidelines for practical implementation of the code case, EPRI sponsored work to develop evaluation procedures and criteria. As part of an EPRI sponsored pilot study, these procedures have been applied to a BWR plant. Piping within the scope of the existing Section XI program has been analyzed. The results of this effort indicate that implementation of RBISI programs can significantly reduce the cost and radiation exposure associated with in-service inspections. The revised program was compared to the previous program and a net gain in safety benefit was demonstrated.

  15. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    SciTech Connect

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  16. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    SciTech Connect

    Orlov, Andrey; Degueldre, Claude; Kaufmann, Wilfried

    2013-01-15

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  17. Accident sequence analysis for a BWR (Boiling Water Reactor) during low power and shutdown operations

    SciTech Connect

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs.

  18. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    SciTech Connect

    Not Available

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  19. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    SciTech Connect

    Not Available

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  20. Radiolysis of the coolant in the VK-50 boiling water reactor

    SciTech Connect

    Zabelin, A.I.; Shmelev, V.E.

    1986-10-01

    Radiolysis of the coolant proceeds at a higher rate in a boiling water reactor as compared to a water-moderated, water-cooled reactor. The radiolytic gases (hydrogen and oxygen) exiting the reactor together with steam can form a potentially explosive mixture. Special interest attaches to the results obtained under the codnitions of prolonged operation of the VK-50 reactor. Tests of various water-chemistry conditions which were performed in the experimental reactor showed their critical influence on the rate of progress of radiolytic processes. The entire period of operation of the reactor may be arbitrarily divided into three stages, each of which is characterized by its own peculiar conditions of water chemistry and range of thermal power. From stage to stage, there is a noticeable improvement in the coolant quality which to a limited extent is reflected in the exit of radiolytic gases with the steam. The concentration of radiolytic gases increases with decreased power and with an increased content of corrosion products and other contaminants in the coolant.

  1. Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

    SciTech Connect

    Serebrov, A. P. Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; and others

    2015-10-15

    In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

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

    SciTech Connect

    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.

  3. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  4. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  5. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, Paul R.

    1994-01-01

    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  6. Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor

    SciTech Connect

    Diercks, D.R.; Gaitonde, S.M.

    1982-09-01

    The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.

  7. Thermalhydraulic processes in the reactor coolant system of a BWR (boiling water reactor) under severe accident conditions

    SciTech Connect

    Hodge, S.A.

    1989-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MSIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a brief discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermalhydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are briefly described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom Atomic Power Station is presented. 12 refs., 4 figs., 1 tab.

  8. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    SciTech Connect

    Korteniemi, V.; Haapalehto, T.; Puustinen, M.

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  9. Boiling Heat Transfer Experiments by using Transparent Heated Microtube

    NASA Astrophysics Data System (ADS)

    Huang, Shih-Che; Kawanami, Osamu; Kawakami, Kazunari; Honda, Itsuro; Kawashima, Yousuke; Ohta, Haruhiko

    For detailed study of the relationship between boiling bubble behavior and inner wall temperature during flow boiling in microtubes, a transparent heated microtube, whose inner wall was coated with a thin gold film, was employed. Boiling behavior could be observed clearly, and the inner wall temperature of the tube was measured simultaneously with direct heating of the film. Ionized water was used as a test fluid. The experimental conditions were as follows: tube diameter, 1 mm; inlet liquid subcooling, 10 K; mass velocity, 100 kg/m2s and heat flux, up to 469 kW/m2 in the open system. As a result, the frequencies of fluctuation of the inner wall temperature and flow rate were divided into four regions. In addition, the fluctuation range of flow rate increased with increasing heat flux however, this fluctuation decreased drastically for heat flux over 212 kW/m2. The fluctuation of void fraction coincided with that of inner wall temperature.

  10. Implementation of a source term control program in a mature boiling water reactor.

    PubMed

    Vargo, G J; Jarvis, A J; Remark, J F

    1991-06-01

    The implementation and results of a source term control program implemented at the James A. FitzPatrick Nuclear Power Plant (JAF), a mature boiling water reactor (BWR) facility that has been in commercial operation since 1975, are discussed. Following a chemical decontamination of the reactor water recirculation piping in the Reload 8/Cycle 9 refueling outage in 1988, hydrogen water chemistry (HWC) and feedwater Zn addition were implemented. This is the first application of both HWC and feedwater Zn addition in a BWR facility. The radiological benefits and impacts of combined operation of HWC and feedwater Zn addition at JAF during Cycle 9 are detailed and summarized. The implementation of hydrogen water chemistry resulted in a significant transport of corrosion products within the reactor coolant system that was greater than anticipated. Feedwater Zn addition appears to be effective in controlling buildup of other activated corrosion products such as 60Co on reactor water recirculation piping; however, adverse impacts were encountered. The major adverse impact of feedwater Zn addition is the production of 65Zn that is released during plant outages and operational transients. PMID:2032839

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

    SciTech Connect

    Not Available

    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.

  12. Non Invasive Water Level Monitoring on Boiling Water Reactors Using Internal Gamma Radiation: Application of Soft Computing Methods

    SciTech Connect

    Fleischer, Sebastian; Hampel, Rainer

    2006-07-01

    To provide best knowledge about safety-related water level values in boiling water reactors (BWR) is essentially for operational regime. For the water level determination hydrostatic level measurement systems are almost exclusively applied, because they stand the test over many decades in conventional and nuclear power plants (NPP). Due to the steam generation especially in BWR a specific phenomenon occurs which leads to a water-steam mixture level in the reactor annular space and reactor plenum. The mixture level is a high transient non-measurable value concerning the hydrostatic water level measuring system and it significantly differs from the measured collapsed water level. In particular, during operational and accidental transient processes like fast negative pressure transients, the monitoring of these water levels is very important. In addition to the hydrostatic water level measurement system a diverse water level measurement system for BWR should be used. A real physical diversity is given by gamma radiation distribution inside and outside the reactor pressure vessel correlating with the water level. The vertical gamma radiation distribution depends on the water level, but it is also a function of the neutron flux and the coolant recirculation pump speed. For the water level monitoring, special algorithms are required. An analytical determination of the gamma radiation distribution outside the reactor pressure vessel is impossible due to the multitude of radiation of physical processes, complicated non-stationary radiation source distribution and complex geometry of fixtures. For creating suited algorithms Soft Computing methods (Fuzzy Sets Theory, Artificial Neural Networks, etc.) will be used. Therefore, a database containing input values (gamma radiation distribution) and output values (water levels) had to be built. Here, the database was established by experiments (data from BWR and from a test setup) and simulation with the authorised thermo

  13. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    SciTech Connect

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  14. Review of Reactor Neutrino Experiments

    NASA Astrophysics Data System (ADS)

    Kim, Soo-Bong

    New generation of reactor neutrino experiments, Daya Bay and RENO, have made definitive measurements of the smallest neutrino mixing angle θ13 in 2012, based on the disappearance of electron antineutrinos. More precise measurements of the mixing angle and reactor neutrino spectra have been made and presented. A rather large value of θ13 has opened a new window to find the CP violation phase and to determine the neutrino mass hierarchy. Future reactor experiments, JUNO and RENO-50, are proposed to determine the neutrino mass hierarchy and to make highly precise measurements of θ12, Δm212, and Δm312.

  15. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR... appendix. B. Generic technical specifications means the information, required by 10 CFR 50.36 and 50.36a... for the intended application. H. All other terms in this appendix have the meaning set out in 10...

  16. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR... appendix. B. Generic technical specifications means the information, required by 10 CFR 50.36 and 50.36a... for the intended application. H. All other terms in this appendix have the meaning set out in 10...

  17. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-14

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos. (As Shown In Attachment 1), License Nos. (As Shown In Attachment 1), EA-13-109; Order Modifying Licenses With Regard to Reliable Hardened...

  18. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    SciTech Connect

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  19. Experimental determination of residual stress by neutron diffraction in a boiling water reactor core shroud

    SciTech Connect

    Payzant, A.; Spooner, S.; Zhu, Xiaojing; Hubbard, C.R.

    1996-06-01

    Residual strains in a 51 mm (2-inch) thick 304L stainless steel plate have been measured by neutron diffraction and interpreted in terms of residual stress. The plate, measuring (300 mm) in area, was removed from a 6m (20-ft.) diameter unirradiated boiling water reactor core shroud, and included a multiple-pass horizontal weld which joined two of the cylindrical shells which comprise the core shroud. Residual stress mapping was undertaken in the heat affected zone, concentrating on the outside half of the plate thickness. Variations in residual stresses with location appeared consistent with trends expected from finite element calculations, considering that a large fraction of the residual hoop stress was released upon removal of the plate from the core shroud cylinder.

  20. Validation of CASMO/SIMULATE code package for TVO boiling water reactors

    SciTech Connect

    Roine, T.; Anttila, M.; Hoeglund, R.; Solala, M.

    1995-12-31

    The possibilities of modelling TVO I and TVO II boiling water reactors with the CASMO/SIMULATE code package have been studied by VTT Energy. Altogether 28 cycles have been calculated, concentrating on a few cycles with both equilibrium and mixed fuel loadings. The comparisons have been performed mainly against plant measured data (TIP measurements), but also against results of core follow calculations with other programs. Special emphasis has been put on variation of the effective multiplication factor (k{sub eff}), both in a cycle and between cycles. With a minimum of effort a large data set has been produced and the capability to reliably and easily use the CASMO/SIMULATE code has been proven. In addition a good basis for the refinement of the calculation model has been created.

  1. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  2. Design and test of a compact optics system for the pool boiling experiment

    NASA Technical Reports Server (NTRS)

    Ling, Jerri S.; Laubenthal, James R.

    1990-01-01

    The experiment described seeks to improve the understanding of the fundamental mechanisms that constitute nucleate pool boiling. The vehicle for accomplishing this is an investigation, including tests to be conducted in microgravity and coupled with appropriate analyses, of the heat transfer and vapor bubble dynamics associated with nucleation, bubble growth/collapse and subsequent motion, considering the interrelations between buoyancy, momentum and surface tension which will govern the motion of the vapor and surrounding liquid, as a function of the heating rate at the heat transfer surface and the temperature level and distribution in the bulk liquid. The experiment is designed to be contained within the confines of a Get-Away-Special Canister (GAS Can) installed in the bay of the space shuttle. When the shuttle reaches orbit, the experiment will be turned on and testing will proceed automatically. In the proposed Pool Boiling Experiment a pool of liquid, initially at a precisely defined pressure and temperature, will be subjected to a step imposed heat flux from a semitransparent thin-film heater forming part of one wall of the container such that boiling is initiated and maintained for a defined period of time at a constant pressure level. Transient measurements of the heater surface and fluid temperatures near the surface will be made, noting especially the conditions at the onset of boiling, along with motion photography of the boiling process in two simultaneous views, from beneath the heating surface and from the side. The conduct of the experiment and the data acquisition will be completely automated and self-contained. For the initial flight, a total of nine tests are proposed, with three levels of heat flux and three levels of subcooling. The design process used in the development and check-out of the compact photographic/optics system for the Pool Boiling Experiment is documented.

  3. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    SciTech Connect

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

  4. Neutrino Experiments at Reactors

    DOE R&D Accomplishments Database

    Reines, F.; Gurr, H. S.; Jenkins, T. L.; Munsee, J. H.

    1968-09-09

    A description is given of the electron-antineutrino program using a large fission reactor. A search has been made for a neutral weak interaction via the reaction (electron antineutrino + d .> p + n + electron antineutrino), the reaction (electron antineutrino + d .> n + n + e{sup +}) has now been detected, and an effort is underway to observe the elastic scattering reaction (electron antineutrino + e{sup -} .> electron antineutrino + e{sup -}) as well as to measure more precisely the reaction (electron antineutrino + p .> n + e{sup+}). The upper limit on the elastic scattering reaction which we have obtained with our large composite NaI, plastic, liquid scintillation detector is now about 50 times the predicted value.

  5. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    SciTech Connect

    Not Available

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

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

    SciTech Connect

    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. Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

    NASA Astrophysics Data System (ADS)

    Yoshida, Hiroyuki; Takase, Kazuyuki

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low.

  8. Radwaste generation survey update: Volume 1, Boiling water reactors: Final report

    SciTech Connect

    Daloisio, G.S.; Deltete, C.P.

    1988-02-01

    The Electric Power Research Institute (EPRI) commissioned an operations-related project (RP1557-26) in mid-1986 to update the project data base developed for EPRI Report NP-3370, ''Identification of Radwaste Sources and Reduction Techniques,'' which was published in January 1984. An update was deemed particularly desirable in order to assess the impact on power reactor low level radioactive waste generation of 10 CFR 61, the recent implementation of the 1985 Amendment to The Low Level Waste Policy Act of 1980 (and its potential effects on accelerated waste shipment programs), and the efforts of several plants to implement waste minimization programs over the past several years. These events, as reflected in waste generation rates from 1982 through 1986, should help NP-3370 continue to be a useful document for a plant's radwaste manager in the future. Furthermore, the trends of the past several years presented herein should help to more accurately define utility waste source terms for use in planning on-site storage and developing regional burial facilities. A new data base was developed that includes 1982 through 1986 information, as well as pertinent portions of the 1978 through 1981 data base. The result of the project is a two volume report comprising radwaste related information from more than 95% of the nuclear power plants in commercial operation as of 1986. Volume 1 contains all information pertaining to boiling water reactors (BWRs), while Volume 2 contains information for pressurized water reactors (PWRs). The computerized data base of waste volumes, sources and characteristics for each plant type (BWR or PWR) is included as an appendix in each respective volume.

  9. A bifurcation analysis of boiling water reactor on large domain of parametric spaces

    NASA Astrophysics Data System (ADS)

    Pandey, Vikas; Singh, Suneet

    2016-09-01

    The boiling water reactors (BWRs) are inherently nonlinear physical system, as any other physical system. The reactivity feedback, which is caused by both moderator density and temperature, allows several effects reflecting the nonlinear behavior of the system. Stability analyses of BWR is done with a simplified, reduced order model, which couples point reactor kinetics with thermal hydraulics of the reactor core. The linear stability analysis of the BWR for steady states shows that at a critical value of bifurcation parameter (i.e. feedback gain), Hopf bifurcation occurs. These stable and unstable domains of parametric spaces cannot be predicted by linear stability analysis because the stability of system does not include only stability of the steady states. The stability of other dynamics of the system such as limit cycles must be included in study of stability. The nonlinear stability analysis (i.e. bifurcation analysis) becomes an indispensable component of stability analysis in this scenario. Hopf bifurcation, which occur with one free parameter, is studied here and it formulates birth of limit cycles. The excitation of these limit cycles makes the system bistable in the case of subcritical bifurcation whereas stable limit cycles continues in an unstable region for supercritical bifurcation. The distinction between subcritical and supercritical Hopf is done by two parameter analysis (i.e. codimension-2 bifurcation). In this scenario, Generalized Hopf bifurcation (GH) takes place, which separates sub and supercritical Hopf bifurcation. The various types of bifurcation such as limit point bifurcation of limit cycle (LPC), period doubling bifurcation of limit cycles (PD) and Neimark-Sacker bifurcation of limit cycles (NS) have been identified with the Floquet multipliers. The LPC manifests itself as the region of bistability whereas chaotic region exist because of cascading of PD. This region of bistability and chaotic solutions are drawn on the various

  10. Performance of boiling water reactor fuel lead test assemblies to 35 MWd/kg U

    SciTech Connect

    Rowland, T.C.; Ikemoto, R.N.; Gehl, S.

    1986-01-01

    This joint Electric Power Research Institute/General Electric (EPRI/GE) fuel performance program involved thorough preirradiation characterization of fuel used in lead test assemblies (LTAs), detailed surveillance of their operation, and interim site examinations of the assemblies during reactor outages. The program originally included four GE-5 LTAs operating in the Peach Bottom-2 (PB-2) reactor. The program was later modified to include the pressurized fuel rod test assembly in the Peach Bottom-3 (PB-3) reactor. The program modification also included extending the operation of the PB-2 and PB-3 LTA fuel beyond normal discharge exposures. Results are summarized in the paper.

  11. Proposal of experimental setup on boiling two-phase flow on-orbit experiments onboard Japanese experiment module "KIBO"

    NASA Astrophysics Data System (ADS)

    Baba, S.; Sakai, T.; Sawada, K.; Kubota, C.; Wada, Y.; Shinmoto, Y.; Ohta, H.; Asano, H.; Kawanami, O.; Suzuki, K.; Imai, R.; Kawasaki, H.; Fujii, K.; Takayanagi, M.; Yoda, S.

    2011-12-01

    Boiling is one of the efficient modes of heat transfer due to phase change, and is regarded as promising means to be applied for the thermal management systems handling a large amount of waste heat under high heat flux. However, gravity effects on the two-phase flow phenomena and corresponding heat transfer characteristics have not been clarified in detail. The experiments onboard Japanese Experiment Module "KIBO" in International Space Station on boiling two-phase flow under microgravity conditions are proposed to clarify both of heat transfer and flow characteristics under microgravity conditions. To verify the feasibility of ISS experiments on boiling two-phase flow, the Bread Board Model is assembled and its performance and the function of components installed in a test loop are examined.

  12. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    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.

  13. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    SciTech Connect

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  14. Data Reconciliation in the Steam-Turbine Cycle of a Boiling Water Reactor

    SciTech Connect

    Sunde, Svein; Berg, Oivind; Dahlberg, Lennart; Fridqvist, Nils-Olof

    2003-08-15

    A mathematical model for a boiling water reactor steam-turbine cycle was assembled by means of a configurable, steady-state modeling tool TEMPO. The model was connected to live plant data and intermittently fitted to these by minimization of a weighted least-squares object function. The improvement in precision achieved by this reconciliation was assessed from quantities calculated from the model equations linearized around the minimum and from Monte Carlo simulations. It was found that the inclusion of the flow-passing characteristics of the turbines in the model equations significantly improved the precision as compared to simple mass and energy balances, whereas heat transfer calculations in feedwater heaters did not. Under the assumption of linear model equations, the quality of the fit can also be expressed as a goodness-of-fit Q. Typical values for Q were in the order of 0.9. For a validated model Q may be used as a fault detection indicator, and Q dropped to very low values in known cases of disagreement between the model and the plant state. The sensitivity of Q toward measurement faults is discussed in relation to redundancy. The results of the linearized theory and Monte Carlo simulations differed somewhat, and if a more accurate analysis is required, this is better based on the latter. In practical application of the presently employed techniques, however, assessment of uncertainties in raw data is an important prerequisite.

  15. Many-Group Cross-Section Adjustment Techniques for Boiling Water Reactor Adaptive Simulation

    SciTech Connect

    Jessee, Matthew Anderson

    2011-01-01

    Computational capability has been developed to adjust multigroup neutron cross sections, including self-shielding correction factors, to improve the fidelity of boiling water reactor (BWR) core modeling and simulation. The method involves propagating multigroup neutron cross-section uncertainties through various BWR computational models to evaluate uncertainties in key core attributes such as core k{sub eff}, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multigroup cross sections to minimize the disagreement between BWR core modeling predictions and observed (i.e., measured) plant data. For this paper, observed plant data are virtually simulated in the form of perturbed three-dimensional nodal power distributions with the perturbations sized to represent actual discrepancies between predictions and real plant data. The major focus of this work is to efficiently propagate multigroup neutron cross-section uncertainty through BWR lattice physics and core simulator calculations. The data adjustment equations are developed using a subspace approach that exploits the ill-conditioning of the multigroup cross-section covariance matrix to minimize computation and storage burden. Tikhonov regularization is also employed to improve the conditioning of the data adjustment equations. Expressions are also provided for posterior covariance matrices of both the multigroup cross-section and core attributes uncertainties.

  16. Fuzzy logic control of water level in advanced boiling water reactor

    SciTech Connect

    Lin, Chaung; Lee, Chi-Szu; Raghavan, R.; Fahrner, D.M.

    1995-12-31

    The feedwater control system in the Advanced Boiling Water Reactor (ABWR) is more challenging to design compared to other control systems in the plant, due to the possible change in level from void collapses and swells during transient events. A basic fuzzy logic controller is developed using a simplified ABWR mathematical model to demonstrate and compare the performance of this controller with a simplified conventional controller. To reduce the design effort, methods are developed to automatically tune the scaling factors and control rules. As a first step in developing the fuzzy controller, a fuzzy controller with a limited number of rules is developed to respond to normal plant transients such as setpoint changes of plant parameters and load demand changes. Various simulations for setpoint and load demand changes of plant performances were conducted to evaluate the modeled fuzzy logic design against the simplified ABWR model control system. The simulation results show that the performance of the fuzzy logic controller is comparable to that of the Proportional-Integral (PI) controller, However, the fuzzy logic controller produced shorter settling time for step setpoint changes compared to the simplified conventional controller.

  17. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

  18. Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

    SciTech Connect

    Henderson, J M; Wood, S A; Knight, D D

    1981-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting.

  19. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  20. Experimental study of the interaction of pulsations of the neutron flux and the coolant flow in a boiling-water reactor

    SciTech Connect

    Leppik, P.A.

    1984-12-01

    This paper presents results of a study designed to confirm that the interaction of the neutron flux and the coolant flow plays an important role in the mechanism of high-frequency (HF) resonant instability of the VK-50 boiling water reactor. To do this and to check the working model, signals from probes measuring the flow rate of the coolant and the neutron flux were recorded simultaneously (with the help of a magnetograph) in experiments performed in 1981 on driving the VK-50 reactor into the HF reonant instability regimes. Estimates were then obtained for the statistical characteristics of the pulsations of the flow rate and of the neutron flux, including the cross-correlation functions and coherence functions. The basic results of these studies are reported here.

  1. Sustained Heating Experiments for the Investigation of the Boiling Heat Removal at the Upper Surface of Metallic Layer

    SciTech Connect

    Kyoung-Ho Kang; Rae-Joon Park; Sang-Baik Kim; Hee-Dong Kim

    2006-07-01

    Sustained heating experiments are in progress to investigate the boiling configurations and to quantify the heat removal rate at the upper surface of a metallic layer in case of late in-vessel coolant injection. Pure Zinc is used as the simulant of metallic layer and bottom heating via cartridge heater plate is adopted to simulate the heat transfer from the oxidic pool. The test section is rectangular shape having 0.3 m in width and 0.5 m in length. The scoping test was performed. In the scoping test, the height of melt pool is 10 cm and the temperature of the test section bottom was intended to be controlled for maintaining the uniform temperature of 500 deg. C. The scoping test results indicate that the melt pool effectively cooled down via boiling heat removal without energetic melt coolant interaction. The heat removal rate was calculated by 456.27 kW/m{sup 2} {approx} 516.3 kW/m{sup 2} which confirms the effective boiling heat removal in the scoping test. (authors)

  2. Nuclear-coupled thermal-hydraulic stability analysis of boiling water reactors

    NASA Astrophysics Data System (ADS)

    Karve, Atul A.

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model we developed from: the space-time modal neutron kinetics equations based on spatial omega-modes, the equations for two-phase flow in parallel boiling channels, the fuel rod heat conduction equations, and a simple model for the recirculation loop. The model is represented as a dynamical system comprised of time-dependent nonlinear ordinary differential equations, and it is studied using stability analysis, modern bifurcation theory, and numerical simulations. We first determine the stability boundary (SB) 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 and then transform the SB to the practical power-flow map. Using this SB, we show that the normal operating point at 100% power is very stable, stability of points on the 100% rod line decreases as the flow rate is reduced, and that points are least stable in the low-flow/high-power region. We also determine the SB when the modal kinetics is replaced by simple point reactor kinetics and show that the first harmonic mode has no significant effect on the SB. Later we carry out the relevant numerical simulations where we first show that the Hopf bifurcation, that occurs as a parameter is varied across the SB is subcritical, and that, in the important low-flow/high-power region, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line. Hence, a point on the 100% rod line in the low-flow/high-power region, although stable, may nevertheless be a point at which a BWR should not be operated. Numerical simulations are then done to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is determined that the NRC requirement of DR < 0.75-0.8 is not rigorously satisfied in the low

  3. R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

    SciTech Connect

    Vanderhaegen, M.; Paumel, K.; Seiler, J. M.; Tourin, A.; Jeannot, J. P.; Rodriguez, G.

    2011-07-01

    In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boiling intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)

  4. Effect of surface oxidation on the onset of nucleate boiling in a materials test reactor coolant channel

    DOE PAGESBeta

    Forrest, Eric C.; Don, Sarah M.; Hu, Lin -Wen; Buongiorno, Jacopo; McKrell, Thomas J.

    2016-02-29

    The onset of nucleate boiling (ONB) serves as the thermal-hydraulic operating limit for many research and test reactors. However, boiling incipience under forced convection has not been well-characterized in narrow channel geometries or for oxidized surface conditions. This study presents experimental data for the ONB in vertical upflow of deionized (DI) water in a simulated materials test reactor (MTR) coolant channel. The channel gap thickness and aspect ratio were 1.96 mm and 29:1, respectively. Boiling surface conditions were carefully controlled and characterized, with both heavily oxidized and native oxide surfaces tested. Measurements were performed for mass fluxes ranging from 750more » to 3000 kg/m2s and for subcoolings ranging from 10 to 45°C. ONB was identified using a combination of high-speed visual observation, surface temperature measurements, and channel pressure drop measurements. Surface temperature measurements were found to be most reliable in identifying the ONB. For the nominal (native oxide) surface, results indicate that the correlation of Bergles and Rohsenow, when paired with the appropriate single-phase heat transfer correlation, adequately predicts the ONB heat flux. Furthermore, incipience on the oxidized surface occurred at a higher heat flux and superheat than on the plain surface.« less

  5. Study of the oxide layer formed on stainless steel exposed to boiling water reactor conditions by ion beam techniques

    NASA Astrophysics Data System (ADS)

    Degueldre, C.; Buckley, D.; Dran, J. C.; Schenker, E.

    1998-01-01

    The build-up of the oxide layer on austenitic steel under boiling water reactor (BWR) conditions was studied by macro- and micro-Rutherford backscattering spectrometry (RBS) and sputtered neutral mass spectroscopy (SNMS). RBS is applicable when the oxide thickness is larger than 20 nm and yields both the layer thickness and its stoichiometry. SNMS provides elemental depth profiles and the oxide thickness when combined with profilometry. Stainless steel strip samples pre-treated (electro- or mechanically polished) or not, exposed in a loop simulating the BWR-conditions for periods ranging from 31 to 291 days and with a low water flow velocity show oxide layers with a thickness of about 300 to 600 nm. There is no significant increase of the oxide layer thickness after 31 days of exposure. The paper confirms the presence of inner and outer oxide layers and also confirms the stoichiometry M 2O 3 in the external part in contact with the oxygenated water. The oxide layer consists not only of an outer layer and an inner layer but also of a deep apparent oxide/metal interface that is attributed to oxide formation through the steel grain boundaries.

  6. Parametric study of recriticality in a boiling water reactor severe accident

    SciTech Connect

    Shamoun, B.I.; Witt, R.J. . Dept. of Nuclear Engineering and Engineering Physics)

    1994-08-01

    Recriticality is possible in a severe accident if unborated or low boron concentration water is added to a damaged core after control rod melting but before fuel melting. Recriticality in a severe accident in a boiling water reactor was parametrically investigated using the TWODANT code. Eigenvalue calculations for a unit central fuel cell with reflective boundary conditions were performed by solving the two-dimensional multigroup steady-state Boltzman transport equation using TWODANT. Two sets of calculations were performed in this work. The first set of calculations was carried out under three types of normal operating conditions to provide reference values for the accident calculations: (a) cold rodded condition, (b) cold unrodded condition, and (c) hot full-power condition. The eigenvalues at these conditions were found to be 1.055, 1.208, and 1.098, respectively. The second set of calculations was carried out after the melting of the control element and during the reflood phase, under the following reflood conditions: (a) reflood with unborated water and (b) reflood with borated water. For the reflood case with unborated water, five values of void fractions were considered (100, 60, 40, 20, and 0%). Decreasing void fractions represent greater refill levels during the reflood process. The system pressure was taken to be 7 MPa, while the moderator temperature was set to 560 K. Plotting the eigenvalue compared with the fraction of control materials lost indicates recriticality is only possible if nearly 100% of the control material is lost from the core. Eigenvalue calculations were repeated for short- and long-term recovery conditions of the reflood phase corresponding to maximum moderator density at 4 MPa pressure and 525 K moderator temperature and for 1 MPa pressure and 325 K moderator temperature, respectively. Recriticality was again observed to be a concern only after losing 95% ore more of control materials from the unit cell.

  7. Channel Bow in Boiling Water Reactors - Hot Cell Examination Results and Correlation to Measured Bow

    SciTech Connect

    Mahmood, S.T.; Lin, Y.P.; Dubecky, M.A.; Mader, E.V.

    2007-07-01

    An increase in frequency of fuel channel-control blade interference has been observed in Boiling Water Reactors (BWR) in recent years. Many of the channels leading to interference were found to bow towards the control blade in a manner that was inconsistent with the expected bow due to other effects. The pattern of bow appeared to indicate a new channel bow mechanism that differed from the predominant bow mechanism caused by differential growth due to fast-fluence gradients. In order to investigate this new type of channel bow, coupons from several channels with varying degrees of bow were returned to the GE Vallecitos Nuclear Center (VNC) for Post-Irradiation Examination (PIE). This paper describes the characteristics of channel corrosion and hydrogen pickup observed, and relates the observations to the channel exposure level, control history, and measured channel bow. The channels selected for PIE had exposures in the range of 36-48 GWd/MTU and covered a wide range of measured bow. The coupons were obtained at 4 elevations from opposing channel sides adjacent and away from the control blade. The PIE performed on these coupons included visual examination, metallography, and hydrogen concentration measurements. A new mechanism of control-blade shadow corrosion-induced channel bow was found to correlate with differences in the extent of corrosion and corresponding differences in the hydrogen concentration between opposite sides of the channels. The increased corrosion on the control blade sides was found to be dependent on the level of control early in the life of the channel. The contributions of other potential factors leading to increased channel bow and channel-control blade interference are also discussed in this paper. (authors)

  8. The Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Wang, Zhe

    2010-02-01

    The 3x3 PMNS leptonic mixing matrix relates the mass and flavor eigenstates of the 3 known neutrinos. The θ13 mixing angle is the last unknown mixing angle in the PMNS matrix, the parameters of which must be determined experimentally. The Daya Bay experiment will search for the ``disappearance'' of reactor anti-neutrinos from the Daya Bay and Ling Ao Nuclear Power Plants located in Daya Bay, Guangdong, China using multiple identical detectors at different baselines. The disappearance probability of reactor anti-neutrinos at short baselines of 1-2km is directly proportional to 2̂(2 θ13). The goal of the Daya Bay experiment is to reach a sensitivity of 2̂(2 θ13) = 0.01 at the 90 % C.L. The status and prospects of the experiment will be presented. )

  9. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    SciTech Connect

    Choi, Y.A.; Feltus, M.A.

    1995-07-01

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company`s Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates.

  10. Comparing Simulation Results with Traditional PRA Model on a Boiling Water Reactor Station Blackout Case Study

    SciTech Connect

    Zhegang Ma; Diego Mandelli; Curtis Smith

    2011-07-01

    A previous study used RELAP and RAVEN to conduct a boiling water reactor station black-out (SBO) case study in a simulation based environment to show the capabilities of the risk-informed safety margin characterization methodology. This report compares the RELAP/RAVEN simulation results with traditional PRA model results. The RELAP/RAVEN simulation run results were reviewed for their input parameters and output results. The input parameters for each simulation run include various timing information such as diesel generator or offsite power recovery time, Safety Relief Valve stuck open time, High Pressure Core Injection or Reactor Core Isolation Cooling fail to run time, extended core cooling operation time, depressurization delay time, and firewater injection time. The output results include the maximum fuel clad temperature, the outcome, and the simulation end time. A traditional SBO PRA model in this report contains four event trees that are linked together with the transferring feature in SAPHIRE software. Unlike the usual Level 1 PRA quantification process in which only core damage sequences are quantified, this report quantifies all SBO sequences, whether they are core damage sequences or success (i.e., non core damage) sequences, in order to provide a full comparison with the simulation results. Three different approaches were used to solve event tree top events and quantify the SBO sequences: “W” process flag, default process flag without proper adjustment, and default process flag with adjustment to account for the success branch probabilities. Without post-processing, the first two approaches yield incorrect results with a total conditional probability greater than 1.0. The last approach accounts for the success branch probabilities and provides correct conditional sequence probabilities that are to be used for comparison. To better compare the results from the PRA model and the simulation runs, a simplified SBO event tree was developed with only four

  11. Flaw density examinations of a clad boiling water reactor pressure vessel segment

    SciTech Connect

    Cook, K.V.; McClung, R.W.

    1986-01-01

    Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel (nominally 0.7 by 3 m (2 by 10 ft)). This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected.

  12. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    SciTech Connect

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  13. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    SciTech Connect

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  14. Electrical design of Space Shuttle payload G-534: The pool boiling experiment

    NASA Technical Reports Server (NTRS)

    Francisco, David R.

    1993-01-01

    Payload G-534, the Pool Boiling Experiment (PBE), is a Get Away Special (GAS) payload that flew on the Space Shuttle Spacelab Mission J (STS 47) on September 19-21, 1992. This paper will give a brief overall description of the experiment with the main discussion being the electrical design with a detailed description of the power system and interface to the GAS electronics. The batteries used and their interface to the experiment Power Control Unit (PCU) and GAS electronics will be examined. The design philosophy for the PCU will be discussed in detail. The criteria for selection of fuses, relays, power semiconductors, and other electrical components along with grounding and shielding policy for the entire experiment are presented. The intent of this paper is to discuss the use of military tested parts and basic design guidelines to build a quality experiment for minimal additional cost.

  15. Electrical design of Space Shuttle payload G-534: The pool boiling experiment

    NASA Astrophysics Data System (ADS)

    Francisco, David R.

    1993-05-01

    Payload G-534, the Pool Boiling Experiment (PBE), is a Get Away Special (GAS) payload that flew on the Space Shuttle Spacelab Mission J (STS 47) on September 19-21, 1992. This paper will give a brief overall description of the experiment with the main discussion being the electrical design with a detailed description of the power system and interface to the GAS electronics. The batteries used and their interface to the experiment Power Control Unit (PCU) and GAS electronics will be examined. The design philosophy for the PCU will be discussed in detail. The criteria for selection of fuses, relays, power semiconductors, and other electrical components along with grounding and shielding policy for the entire experiment are presented. The intent of this paper is to discuss the use of military tested parts and basic design guidelines to build a quality experiment for minimal additional cost.

  16. Analysis and interpretation of low gravity boiling experiments in the KC-135

    NASA Technical Reports Server (NTRS)

    Cuta, Judith M.; Krotiuk, William J.; Samuels, Jeffery W.

    1988-01-01

    Reduced gravity two phase flow boiling and condensing experiments were conducted in the NASA KC-135. In an attempt to gain a better understanding of two phase flow behavior in reduced gravity, the individual test runs were critically examined to determine their suitability to be taken as representative of two phase flow in microgravity. Selected runs were simulated using the two fluid thermal-hydraulic COBRA/TRAC computer code. The comparisons of these sophisticated codes with the observed flow behavior illustrate the generic shortcomings of current two phase modeling capabilities in application to reduced gravity conditions.

  17. Development of a mechanistic model for forced convection subcooled boiling

    NASA Astrophysics Data System (ADS)

    Shaver, Dillon R.

    The focus of this work is on the formulation, implementation, and testing of a mechanistic model of subcooled boiling. Subcooled boiling is the process of vapor generation on a heated wall when the bulk liquid temperature is still below saturation. This is part of a larger effort by the US DoE's CASL project to apply advanced computational tools to the simulation of light water reactors. To support this effort, the formulation of the dispersed field model is described and a complete model of interfacial forces is formulated. The model has been implemented in the NPHASE-CMFD computer code with a K-epsilon model of turbulence. The interfacial force models are built on extensive work by other authors, and include novel formulations of the turbulent dispersion and lift forces. The complete model of interfacial forces is compared to experiments for adiabatic bubbly flows, including both steady-state and unsteady conditions. The same model is then applied to a transient gas/liquid flow in a complex geometry of fuel channels in a sodium fast reactor. Building on the foundation of the interfacial force model, a mechanistic model of forced-convection subcooled boiling is proposed. This model uses the heat flux partitioning concept and accounts for condensation of bubbles attached to the wall. This allows the model to capture the enhanced heat transfer associated with boiling before the point of net generation of vapor, a phenomenon consistent with existing experimental observations. The model is compared to four different experiments encompassing flows of light water, heavy water, and R12 at different pressures, in cylindrical channels, an internally heated annulus, and a rectangular channel. The experimental data includes axial and radial profiles of both liquid temperature and vapor volume fraction, and the agreement can be considered quite good. The complete model is then applied to simulations of subcooled boiling in nuclear reactor subchannels consistent with the

  18. Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2013-02-01

    A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  19. Experience with non-fuel-bearing components in LWR (light-water reactor) fuel systems

    SciTech Connect

    Bailey, W.J.; Berting, F.M.

    1990-12-01

    Many non-fuel-bearing components are so closely associated with the spent fuel assemblies that their integrity and behavior must be taken into consideration with the fuel assemblies, when handling spent fuel of planning waste management activities. Presented herein is some of the experience that has been gained over the past two decades from non-fuel-bearing components in light-water reactors (LWRs), both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). Among the most important of these components are the control rod systems, the absorber and burnable poison rods, and the fuel assembly channels. 15 refs., 5 figs., 2 tabs.

  20. The Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Bishai, Mary

    2007-04-01

    The 3x3 PMNS leptonic mixing matrix relates the mass and flavor eigenstates of the 3 known neutrinos. The θ13 mixing angle is the last unknown mixing angle in the PMNS matrix, the parameters of which must be determined experimentally. The goal of the Daya Bay experiment is to measure θ13 with a sensitivity in 2circ(2θ13) of 0.01. The Daya Bay experiment will search for the `disappearance' of reactor electron anti-neutrinos from the Daya Bay and Ling Ao Nuclear Power Plants located in Daya Bay, Guangdong, China using multiple identical detectors at different baselines. The status and prospects of the experiment will be presented.

  1. Comparison of Computational Results with a Low-g, Nitrogen Slosh and Boiling Experiment

    NASA Technical Reports Server (NTRS)

    Stewart, Mark E.; Moder, Jeffrey P.

    2015-01-01

    This paper compares a fluid/thermal simulation, in Fluent, with a low-g, nitrogen slosh and boiling experiment. In 2010, the French Space Agency, CNES, performed cryogenic nitrogen experiments in a low-g aircraft campaign. From one parabolic flight, a low-g interval was simulated that focuses on low-g motion of nitrogen liquid and vapor with significant condensation, evaporation, and boiling. The computational results are compared with high-speed video, pressure data, heat transfer, and temperature data from sensors on the axis of the cylindrically shaped tank. These experimental and computational results compare favorably. The initial temperature stratification is in good agreement, and the two-phase fluid motion is qualitatively captured. Temperature data is matched except that the temperature sensors are unable to capture fast temperature transients when the sensors move from wet to dry (liquid to vapor) operation. Pressure evolution is approximately captured, but condensation and evaporation rate modeling and prediction need further theoretical analysis.

  2. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    NASA Astrophysics Data System (ADS)

    Trianti, Nuri; Su'ud, Zaki; Arif, Idam; Riyana, EkaSapta

    2014-09-01

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  3. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    SciTech Connect

    Trianti, Nuri E-mail: szaki@fi.itba.c.id; Su'ud, Zaki E-mail: szaki@fi.itba.c.id; Arif, Idam E-mail: szaki@fi.itba.c.id; Riyana, EkaSapta

    2014-09-30

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  4. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J. ); Rey, J.M. )

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of unexpected'' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a new and improved'' state of the art has emerged recently.

  5. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J.; Rey, J.M.

    1992-05-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of ``unexpected`` instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a ``new and improved`` state of the art has emerged recently.

  6. Pool boiling of water-Al2O3 and water-Cu nanofluids on horizontal smooth tubes.

    PubMed

    Cieslinski, Janusz T; Kaczmarczyk, Tomasz Z

    2011-01-01

    Experimental investigation of heat transfer during pool boiling of two nanofluids, i.e., water-Al2O3 and water-Cu has been carried out. Nanoparticles were tested at the concentration of 0.01%, 0.1%, and 1% by weight. The horizontal smooth copper and stainless steel tubes having 10 mm OD and 0.6 mm wall thickness formed test heater. The experiments have been performed to establish the influence of nanofluids concentration as well as tube surface material on heat transfer characteristics at atmospheric pressure. The results indicate that independent of concentration nanoparticle material (Al2O3 and Cu) has almost no influence on heat transfer coefficient while boiling of water-Al2O3 or water-Cu nanofluids on smooth copper tube. It seems that heater material did not affect the boiling heat transfer in 0.1 wt.% water-Cu nanofluid, nevertheless independent of concentration, distinctly higher heat transfer coefficient was recorded for stainless steel tube than for copper tube for the same heat flux density. PMID:21711741

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

    SciTech Connect

    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.

  8. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    SciTech Connect

    Edwards, Robert; Huang, Zhengyu

    2001-06-17

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown.

  9. Subcooled Pool Boiling Heat Transfer Mechanisms in Microgravity: Terrier-improved Orion Sounding Rocket Experiment

    NASA Technical Reports Server (NTRS)

    Kim, Jungho; Benton, John; Kucner, Robert

    2000-01-01

    A microscale heater array was used to study boiling in earth gravity and microgravity. The heater array consisted of 96 serpentine heaters on a quartz substrate. Each heater was 0.27 square millimeters. Electronic feedback loops kept each heater's temperature at a specified value. The University of Maryland constructed an experiment for the Terrier-Improved Orion sounding rocket that was delivered to NASA Wallops and flown. About 200 s of high quality microgravity and heat transfer data were obtained. The VCR malfunctioned, and no video was acquired. Subsequently, the test package was redesigned to fly on the KC-135 to obtain both data and video. The pressure was held at atmospheric pressure and the bulk temperature was about 20 C. The wall temperature was varied from 85 to 65 C. Results show that gravity has little effect on boiling heat transfer at wall superheats below 25 C, despite vast differences in bubble behavior between gravity levels. In microgravity, a large primary bubble was surrounded by smaller bubbles, which eventually merged with the primary bubble. This bubble was formed by smaller bubbles coalescing, but had a constant size for a given superheat, indicating a balance between evaporation at the base and condensation on the cap. Most of the heaters under the bubble indicated low heat transfer, suggesting dryout at those heaters. High heat transfer occurred at the contact line surrounding the primary bubble. Marangoni convection formed a "jet" of fluid into the bulk fluid that forced the bubble onto the heater.

  10. Uncertainty analysis of fission fraction for reactor antineutrino experiments

    NASA Astrophysics Data System (ADS)

    Ma, X. B.; Lu, F.; Wang, L. Z.; Chen, Y. X.; Zhong, W. L.; An, F. P.

    2016-06-01

    Reactor simulation is an important source of uncertainties for a reactor neutrino experiment. Therefore, how to evaluate the antineutrino flux uncertainty results from reactor simulation is an important question. In this study, a method of the antineutrino flux uncertainty result from reactor simulation was proposed by considering the correlation coefficient. In order to use this method in the Daya Bay antineutrino experiment, the open source code DRAGON was improved and used for obtaining the fission fraction and correlation coefficient. The average fission fraction between DRAGON and SCIENCE code was compared and the difference was less than 5% for all the four isotopes. The uncertainty of fission fraction was evaluated by comparing simulation atomic density of four main isotopes with Takahama-3 experiment measurement. After that, the uncertainty of the antineutrino flux results from reactor simulation was evaluated as 0.6% per core for Daya Bay antineutrino experiment.

  11. The Zero Boil-Off Tank Experiment Contributions to the Development of Cryogenic Fluid Management

    NASA Technical Reports Server (NTRS)

    Chato, David J.; Kassemi, Mohammad

    2015-01-01

    The Zero Boil-Off Technology (ZBOT) Experiment involves performing a small scale ISS experiment to study tank pressurization and pressure control in microgravity. The ZBOT experiment consists of a vacuum jacketed test tank filled with an inert fluorocarbon simulant liquid. Heaters and thermo-electric coolers are used in conjunction with an axial jet mixer flow loop to study a range of thermal conditions within the tank. The objective is to provide a high quality database of low gravity fluid motions and thermal transients which will be used to validate Computational Fluid Dynamic (CFD) modeling. This CFD can then be used in turn to predict behavior in larger systems with cryogens. This paper will discuss the current status of the ZBOT experiment as it approaches its flight to installation on the International Space Station, how its findings can be scaled to larger and more ambitious cryogenic fluid management experiments, as well as ideas for follow-on investigations using ZBOT like hardware to study other aspects of cryogenic fluid management.

  12. A Simple Tubular Reactor Experiment.

    ERIC Educational Resources Information Center

    Hudgins, Robert R.; Cayrol, Bertrand

    1981-01-01

    Using the hydrolysis of crystal violet dye by sodium hydroxide as an example, the theory, apparatus, and procedure for a laboratory demonstration of tubular reactor behavior are described. The reaction presented can occur at room temperature and features a color change to reinforce measured results. (WB)

  13. Boiling eXperiment Facility (BXF) Fluid Toxicity Technical Interchange Meeting (TIM) with the Payload Safety Review Panel (PSRP)

    NASA Technical Reports Server (NTRS)

    Sheredy, William A.

    2012-01-01

    A Technical Interchange meeting was held between the payload developers for the Boiling eXperiment Facility (BXF) and the NASA Safety Review Panel concerning operational anomaly that resulted in overheating one of the fluid heaters, shorted a 24VDC power supply and generated Perfluoroisobutylene (PFiB) from Perfluorohexane.

  14. Analysis of cracked core spray injection line piping from the Quad Cities Units 1 and 2 boiling water reactors

    SciTech Connect

    Diercks, D.R.

    1983-12-01

    Elbow assemblies and adjacent piping from the loops A and B core spray injection lines of Quad Cities Units 1 and 2 Boiling Water Reactors have been examined in order to determine the nature and causes of coolant leakages and flaw indications detected during hydrostatic tests and subsequent ultrasonic inspections. The elbow assemblies were found to contain multiple intergranular cracks in the weld heat-affected zones. The cracking was predominantly axial in orientation in the forged elbow and wedge components, whereas mixed axial and circumferential cracking was seen in the wrought piping pieces. In at least two instances, axial cracks completely penetrated the circumferential weld joining adjacent components. Based upon the observations made in the present study, the failures were attributed to intergranular stress corrosion cracking caused by the weld-induced sensitized microstructure and residual stresses present; dissolved oxygen in the reactor coolant apparently served as the corrosive species. The predominantly axial orientation of the cracks present in the forged components is believed to be related to the banded microstructure present in these components. The metallographic studies reported are supplemented by x-radiography, chemical analysis and mechanical test results, determinations of the degree of sensitization present, and measurements of weld metal delta ferrite content.

  15. Future Reactor Neutrino Experiments (RRNOLD)1

    NASA Astrophysics Data System (ADS)

    Jaffe, David E.

    The prospects for future reactor neutrino experiments that would use tens of kilotons of liquid scintillator with a ∼ 50 km baseline are discussed. These experiments are generically dubbed "RRNOLD" for Radical Reactor Neutrino Oscillation Liquid scintillator Detector experiment. Such experiments are designed to resolve the neutrino mass hierarchy and make sub-percent measurements sin2θ12, Δm232 and Δm122 . RRNOLD would also be sensitive to neutrinos from other sources and have notable sensitivity to proton decay.

  16. Probability of failure in BWR (Boiling Water Reactor) reactor coolant piping: Volume 2, Pipe failure induced by crack growth and failure of intermediate supports

    SciTech Connect

    Lo, T.; Bumpus, S.E.; Chinn, D.J.; Mensing, R.W.; Holman, G.S.

    1989-03-01

    The US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to conduct a study to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in the major coolant piping systems of nuclear power plants is large enough to warrant the current stringent design requirements of designing against the postulated effects of a DEGB. The study includes both the PWR (Pressurized Water Reactor) and the BWR (Boiling Water Reactor) plants in the United States. Following the study of PWR plants, a study of BWR reactor coolant piping was performed. The Brunswick Steam Electric Plant at Southport, North Carolina was selected as the pilot plant for the BWR evaluation. The probability of pipe failure in three major coolant pipings was assessed: the recirculation loops, the primary steam lines, and the main feedwater lines. In the case of recirculation loops, both the existing and a proposed replacement system were studied. A probabilistic fracture mechanics approach was used in this study to estimate the crack growth and to assess the crack stability in the piping systems throughout the lifetime of the plant. The effects of the failure of intermediate pipe supports were also examined. The results of the assessment indicated that the probability of occurrence of DEGB due to crack growth and instability is small if the problem of intergranular stress corrosion cracking (IGSCC) is resolved by the use of the replacement system. 33 refs., 41 figs., 32 tabs.

  17. Environmental qualification testing of payload G-534, the Pool Boiling Experiment

    NASA Technical Reports Server (NTRS)

    Sexton, J. Andrew

    1992-01-01

    Payload G-534, the prototype Pool Boiling Experiment (PBE), is scheduled to fly on the STS-47 mission in September 1992. This paper describes the purpose of the experiment and the environmental qualification testing program that was used to prove the integrity of the hardware. Component and box level vibration and thermal cycling tests were performed to give an early level of confidence in the hardware designs. At the system level, vibration, thermal extreme soaks, and thermal vacuum cycling tests were performed to qualify the complete design for the expected shuttle environment. The system level vibration testing included three axis sine sweeps and random inputs. The system level hot and cold soak tests demonstrated the hardware's capability to operate over a wide range of temperatures and gave wider latitude in determining which shuttle thermal attitudes were compatible with the experiment. The system level thermal vacuum cycling tests demonstrated the hardware's capability to operate in a convection free environment. A unique environmental chamber was designed and fabricated by the PBE team and allowed most of the environmental testing to be performed within the hardware build laboratory. The completion of the test program gave the project team high confidence in the hardware's ability to function as designed during flight.

  18. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    NASA Astrophysics Data System (ADS)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

  19. Secondary pool boiling effects

    NASA Astrophysics Data System (ADS)

    Kruse, C.; Tsubaki, A.; Zuhlke, C.; Anderson, T.; Alexander, D.; Gogos, G.; Ndao, S.

    2016-02-01

    A pool boiling phenomenon referred to as secondary boiling effects is discussed. Based on the experimental trends, a mechanism is proposed that identifies the parameters that lead to this phenomenon. Secondary boiling effects refer to a distinct decrease in the wall superheat temperature near the critical heat flux due to a significant increase in the heat transfer coefficient. Recent pool boiling heat transfer experiments using femtosecond laser processed Inconel, stainless steel, and copper multiscale surfaces consistently displayed secondary boiling effects, which were found to be a result of both temperature drop along the microstructures and nucleation characteristic length scales. The temperature drop is a function of microstructure height and thermal conductivity. An increased microstructure height and a decreased thermal conductivity result in a significant temperature drop along the microstructures. This temperature drop becomes more pronounced at higher heat fluxes and along with the right nucleation characteristic length scales results in a change of the boiling dynamics. Nucleation spreads from the bottom of the microstructure valleys to the top of the microstructures, resulting in a decreased surface superheat with an increasing heat flux. This decrease in the wall superheat at higher heat fluxes is reflected by a "hook back" of the traditional boiling curve and is thus referred to as secondary boiling effects. In addition, a boiling hysteresis during increasing and decreasing heat flux develops due to the secondary boiling effects. This hysteresis further validates the existence of secondary boiling effects.

  20. APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND

    DOEpatents

    Treshow, M.

    1963-07-23

    A method of controlling a fuel-rod-in-tube-type boilingwater reactor having nozzles at the point of water entry into the tube is described. Water is pumped into the nozzles by an auxiliary pump operated by steam from an interstage position of the associated turbine, so that the pumping speed is responsive to turbine demand. (AEC)

  1. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-20

    ... U.S. ABWR design certification (Docket No. 52-001) on December 1, 2009 (74 FR 62829). On June 12, 2009 (74 FR 28112), the NRC amended its regulations to require applicants for new nuclear power reactor... Design Certification VIII. Plain Language IX. Voluntary Consensus Standards X. Finding of No...

  2. Tritium experience in the Tokamak Fusion Test Reactor

    SciTech Connect

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Brooks, J.N.; Hogan, J.

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  3. Optimization of boiling water reactor control rod patterns using linear search

    SciTech Connect

    Kiguchi, T.; Doi, K.; Fikuzaki, T.; Frogner, B.; Lin, C.; Long, A.B.

    1984-10-01

    A computer program for searching the optimal control rod pattern has been developed. The program is able to find a control rod pattern where the resulting power distribution is optimal in the sense that it is the closest to the desired power distribution, and it satisfies all operational constraints. The search procedure consists of iterative uses of two steps: sensitivity analyses of local power and thermal margins using a three-dimensional reactor simulator for a simplified prediction model; linear search for the optimal control rod pattern with the simplified model. The optimal control rod pattern is found along the direction where the performance index gradient is the steepest. This program has been verified to find the optimal control rod pattern through simulations using operational data from the Oyster Creek Reactor.

  4. Flow Boiling and Condensation Experiment (FBCE) for the International Space Station

    NASA Technical Reports Server (NTRS)

    Mudawar, Issam; Hasan, Mohammad M.; Kharangate, Chirag; O'Neill, Lucas; Konishi, Chris; Nahra, Henry; Hall, Nancy; Balasubramaniam, R.; Mackey, Jeffrey

    2015-01-01

    The proposed research aims to develop an integrated two-phase flow boiling/condensation facility for the International Space Station (ISS) to serve as primary platform for obtaining two-phase flow and heat transfer data in microgravity.

  5. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    SciTech Connect

    1997-05-01

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

  6. Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors

    SciTech Connect

    Dykin, V.; Pazsit, I.

    2012-07-01

    A possibility to reconstruct the axial void profile from the simulated in-core neutron noise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heated channel is considered. For this purpose, a self-contained model of the two-phase flow regime is constructed which has quantitatively and qualitatively similar properties to those observed in real BWRs. The model is subsequently used to simulate the signals of neutron detectors induced by the corresponding perturbations in the flow density. The bubbles are generated randomly in both space and time using Monte-Carlo techniques. The axial distribution of the bubble production is chosen such that the mean axial void fraction and void velocity follow the actual values of BWRs. The induced neutron noise signals are calculated and then processed by the standard signal analysis methods such as Auto-Power Spectral Density (APSD) and Cross-Power Spectral Density (CPSD). Two methods for axial void and velocity profiles reconstruction are discussed: the first one is based on the change of the break frequency of the neutron auto-power spectrum with axial core elevation, while the second refers to the estimation of transit times of propagating steam fluctuations between different axial detector positions. This paper summarizes the principles of the model and presents a numerical testing of the qualitative applicability to estimate the required parameters for the reconstruction of the void fraction profile from the neutron noise measurements. (authors)

  7. BWR (boiling-water reactor) radiation control: In-plant demonstration at Vermont Yankee: Final report

    SciTech Connect

    Palino, G.F.; Hobart, R.L.; Sawochka, S.G.

    1987-10-01

    Results of the RP1934 program, which was established by EPRI in 1981 to demonstrate the adequacy of BRAC program (RP819) principles for BWR radiation control at Vermont Yankee, are presented. Evaluations were performed of the effectiveness of optimization of purification system performance, control of feedwater dissolved oxygen concentrations, minimization of corrosion product and ionic transport, and improved startup, shutdown, and layup practices. The impact on shutdown radiation levels of these corrective actions was assessed based on extensive primary system radiation survey and component gamma scan data. Implementation of the BRAC recommendations was found to be insufficient to reduce the rate of activity buildup on out-of-core surfaces at Vermont Yankee, and additional corrective actions were found necessary. Specifically, replacement of cobalt-bearing materials in the control rod drive pins and rollers and feedwater regulating valves was pursued as was installation of electropolished 316 stainless steel during a recirculation piping replacement program. Aggressive programs to further reduce copper concentrations in the reactor water by improving condensate demineralizer efficiency and to minimize organic ingress to the power cycle by reducing organic concentrations in recycled radwaste also were undertaken. Evaluations of the impact on activity buildup of several pretreatment processes including prefilming in moist air, preexposure to high temperature water containing zinc, and electropolishing also were performed in a test loop installed in the reactor water cleanup system. A significant beneficial impact of electropolishing was shown to be present for periods up to 6000 hours.

  8. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Roudén, Jenny; Green, Eva-Lena

    2016-02-01

    This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ˜ 25 effective full power years) of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV) fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M) and calculated (C) results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE)/C ratios of 1.10 for both neutron (E >1.0 MeV) flux and iron atom displacement rate.

  9. Fast critical experiment data for space reactors

    NASA Astrophysics Data System (ADS)

    Collins, P. J.; McFarlane, H. F.; Olsen, D. N.; Atkinson, C. A.; Ross, J. R.

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, about 20 fast benchmark critical experiments were calculated with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors.

  10. Lead chloride crystal growth from boiling solutions

    NASA Astrophysics Data System (ADS)

    Veintemillas-Verdaguer, S.; Rodríguez-Clemente, R.; Torrent-Burgues, J.

    1993-03-01

    Lead chloride single crystals can be grown from boiling solutions using KNO3-H20 solutions as a solvent. Crystals of 1 mm size produced by gel-growth technique were used as seeds. The solubility of PbC12 increases almost linearly with the KNO3 molality being 0.63m in a 7m KNO3 aqueous solutions at 105°C and pH = 2.6; this increase is related to the decrease of the activity coefficient of lead chloride in these solutions. In the first experiments, the supersaturation was attained by solvent extraction, but due to the simultaneous changes in the concentration of the KNO3 mineralizer during the extraction, the growth rate was irregular and defective crystals were obtained. The experimental set-up was therefore modified and a transport technique was added to the system in order to feed the boiling reactor continuously with fresh lead chloride solution. The growth of the crystals takes place at constant concentration of KNO3 in these new conditions. With this experimental modification, isometric PbCI2 crystals of up to lcm size were obtained in three weeks. The observed morphology is close to that calculated by Woensdregt and Hartmann [J. Crystal Growth 87(1988)561].

  11. Environmental Assessment for Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Rincon, Puerto Rico

    SciTech Connect

    N /A

    2003-02-24

    The U.S. Department of Energy (DOE) proposes to consent to a proposal by the Puerto Rico Electric Power Authority (PREPA) to allow public access to the Boiling Nuclear Superheat (BONUS) reactor building located near Rincon, Puerto Rico for use as a museum. PREPA, the owner of the BONUS facility, has determined that the historical significance of this facility, as one of only two reactors of this design ever constructed in the world, warrants preservation in a museum, and that this museum would provide economic benefits to the local community through increased tourism. Therefore, PREPA is proposing development of the BONUS facility as a museum.

  12. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOEpatents

    Oosterkamp, W.J.; Marquino, W.

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

  13. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOEpatents

    Oosterkamp, Willem Jan; Marquino, Wayne

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

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

    SciTech Connect

    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.

  15. US graphite reactor D&D experience

    SciTech Connect

    Garrett, S.M.K.; Williams, N.C.

    1997-02-01

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).

  16. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOEpatents

    Fife, Alex Blair; Ballas, Gary J.

    1998-01-01

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening.

  17. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOEpatents

    Fife, A.B.; Ballas, G.J.

    1998-02-24

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening. 5 figs.

  18. Flow-Boiling Critical Heat Flux Experiments Performed in Reduced Gravity

    NASA Technical Reports Server (NTRS)

    Hasan, Mohammad M.; Mudawar, Issam

    2005-01-01

    Poor understanding of flow boiling in microgravity has recently emerged as a key obstacle to the development of many types of power generation and advanced life support systems intended for space exploration. The critical heat flux (CHF) is perhaps the most important thermal design parameter for boiling systems involving both heatflux-controlled devices and intense heat removal. Exceeding the CHF limit can lead to permanent damage, including physical burnout of the heat-dissipating device. The importance of the CHF limit creates an urgent need to develop predictive design tools to ensure both the safe and reliable operation of a two-phase thermal management system under the reduced-gravity (like that on the Moon and Mars) and microgravity environments of space. At present, very limited information is available on flow-boiling heat transfer and the CHF under these conditions.

  19. Reactor neutrino experiments: θ13 and beyond

    NASA Astrophysics Data System (ADS)

    Qian, Xin; Wang, Wei

    2014-05-01

    We review the current-generation short-baseline reactor neutrino experiments that have firmly established the third neutrino mixing angle θ13 to be nonzero. The relative large value of θ13 (around 9°) has opened many new and exciting opportunities for future neutrino experiments. Daya Bay experiment with the first measurement of Δ m2ee is aiming for a precision measurement of this atmospheric mass-squared splitting with a comparable precision as Δ m2μ μ from accelerator muon neutrino experiments. JUNO, a next-generation reactor neutrino experiment, is targeting to determine the neutrino mass hierarchy (MH) with medium baselines ( 50 km). Beside these opportunities enabled by the large θ13, the current-generation (Daya Bay, Double Chooz, and RENO) and the next-generation (JUNO, RENO-50, and PROSPECT) reactor experiments, with their unprecedented statistics, are also leading the precision era of the three-flavor neutrino oscillation physics as well as constraining new physics beyond the neutrino Standard Model.

  20. PROSPECT: the Precision Reactor Oscillation and Spectrum Experiment

    NASA Astrophysics Data System (ADS)

    Jones, Donald; Prospect Collaboration

    2016-03-01

    PROSPECT is a phased experiment consisting of segmented Li-loaded liquid scintillator antineutrino detectors designed to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. The experiment will be located at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. The first phase is a movable 3 tonne detector located 7-9 m from the compact, highly enriched uranium core. Over the past three years, PROSPECT has deployed multiple prototype detectors at HFIR to understand the local background environment and demonstrate active and passive background rejection. A two-segment prototype has been developed that demonstrates critical subsystems of the full detector. Measuring the neutrino spectrum from 235U will give insight to the recent spectral discrepancies and provide an important benchmark for future reactor experiments. As a high statistics experiment, PROSPECT will probe the sterile neutrino best-fit region within one year of operation at HFIR.

  1. Boils (Furunculosis)

    MedlinePlus

    ... resulting from the deep infection of a hair follicle. The infection is usually caused by a type ... germ gain entry into and infect the hair follicle, resulting in a boil. Boils may resolve with ...

  2. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    SciTech Connect

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two 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 impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  3. Sodium Reactor Experiment decommissioning. Final report

    SciTech Connect

    Carroll, J.W.; Conners, C.C.; Harris, J.M.; Marzec, J.M.; Ureda, B.F.

    1983-08-15

    The Sodium Reactor Experiment (SRE) located at the Rockwell International Field Laboratories northwest of Los Angeles was developed to demonstrate a sodium-cooled, graphite-moderated reactor for civilian use. The reactor reached full power in May 1958 and provided 37 GWh to the Southern California Edison Company grid before it was shut down in 1967. Decommissioning of the SRE began in 1974 with the objective of removing all significant radioactivity from the site and releasing the facility for unrestricted use. Planning documentation was prepared to describe in detail the equipment and techniques development and the decommissioning work scope. A plasma-arc manipulator was developed for remotely dissecting the highly radioactive reactor vessels. Other important developments included techniques for using explosives to cut reactor vessel internal piping, clamps, and brackets; decontaminating porous concrete surfaces; and disposing of massive equipment and structures. The documentation defined the decommissioning in an SRE dismantling plan, in activity requirements for elements of the decommissioning work scope, and in detailed procedures for each major task.

  4. Packaging and shipment of U. S. breeder reactor experiments

    SciTech Connect

    Berger, J.D.

    1980-01-01

    Irradiation testing of fuels and materials in the Fast Test Reactor (FTR) required development of a shipping cask (designated T-3) and associated hardware for loading and shipping of these experiments to postirradiation examination facilities. The T-3 shipping-cask program included design, fabrication, and testing of internal cask packages to protect the experiments during loading, shipping, and unloading. The cask was designed for loading in both the vertical and horizontal attitudes.

  5. Latest progress from the Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    Wang, Zhe; Daya Bay Collaboration

    2016-05-01

    Recently the Daya Bay reactor neutrino experiment has presented several new results about neutrino and reactor physics after acquiring a large data sample and after gaining a more sophisticated understanding of the experiment. In this talk I will introduce the latest progress made by the experiment including a three-flavor neutrino oscillation analysis using neutron capture on gadolinium, which gave sin2 2θ 13 = 0.084 ± 0.005 and |Δm2 ee| = (2.42 ±0.11) × 10-3 eV2, an independent θ 13 measurement using neutron capture on hydrogen, a search for a light sterile neutrino, and a measurement of the reactor antineutrino flux and spectrum.

  6. Assessment of RELAP5/MOD3.1 for gravity-driven injection experiment in the core makeup tank of the CARR Passive Reactor (CP-1300)

    SciTech Connect

    Lee, S.I.; No, H.C.; Bang, Y.S.; Kim, H.J.

    1996-10-01

    The objective of the present work is to improve the analysis capability of RELAP5/MOD3.1 on the direct contact condensation in the core makeup tank (CMT) of passive high-pressure injection system (PHPIS) in the CARR Passive Reactor (CP-1300). The gravity-driven injection experiment is conducted by using a small scale test facility to identify the parameters having significant effects on the gravity-driven injection and the major condensation modes. It turns out that the larger the water subcooling is, the more initiation of injection is delayed, and the sparger and the natural circulation of the hot water from the steam generator accelerate the gravity-driven injection. The condensation modes are divided into three modes: sonic jet, subsonic jet, and steam cavity. RELAP5/MOD3.1 is chosen to evaluate the cod predictability on the direct contact condensation in the CMT. It is found that the predictions of MOD3.1 are in better agreement with the experimental data than those of MOD3.0. From the nodalization study of the test section, the 1-node model shows better agreement with the experimental data than the multi-node models. RELAP5/MOD3.1 identifies the flow regime of the test section as vertical stratification. However, the flow regime observed in the experiment is the subsonic jet with the bubble having the vertical cone shape. To accurately predict the direct contact condensation in the CMT with RELAP5/MOD3.1, it is essential that a new set of the interfacial heat transfer coefficients and a new flow regime map for direct contact condensation in the CMT be developed.

  7. A fast shutdown system for SRS (Savannah River Site) reactors

    SciTech Connect

    Baumann, N.P.

    1990-01-01

    Power has been sharply reduced at Savannah River Site (SRS) reactors in large part to ensure that no bulk boiling occurs during hypothesized loss of coolant accidents. A fast shutdown system is essential to regain much of this lost power. Computations and experiments indicate that a He-3 injection system will serve this function. Instrumented tests of a full system are planned for early 1991 for one of the SRS reactors. 4 refs., 7 figs., 1 tab.

  8. Source term attenuation by water in the Mark I boiling water reactor drywell

    SciTech Connect

    Powers, D.A.

    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.

  9. Status of the RENO Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Joo, K. K.; RENO Collaboration

    2012-08-01

    The Reactor Experiment for Neutrino Oscillation (RENO) is a reactor based neutrino oscillation experiment to measure the neutrino mixing angle θ13 using anti-neutrinos emitted from the Yonggwang nuclear power plant in Korea. Its thermal power output is 16.4 Gwth. The experimental setup consists of two identical 16-ton gadolinium-loaded liquid scintillator detectors. The near and far detectors are placed roughly 290 m and 1.4 km from the center of the reactor array, respectively. The near detector is constructed at underground of a 70 m high hill and the far detector at underground of a 260 m high mountain. The construction of experimental halls and access tunnels for both near and far detector sites was completed in early 2009. The experiment is planned to start data-taking from early 2011. An expected number of observed anti-neutrinos is roughly 1300 per day and 100 per day in the far and near detector, respectively. We expect that an estimated systematic uncertainty is less than 0.5%. With three years of data taken, RENO is sensitive to measure sin2(2θ13)>0.02. Sensitivity is ten times better than the current limit obtained by Chooz. In this review, current status of RENO is presented.

  10. The Corrosion Behavior of Ni3(Si,Nb) Alloys in Boiling 70 wt.% Sulfuric Acid

    NASA Astrophysics Data System (ADS)

    Hsu, Jen-Hsien; Larson, Christopher M.; Newkirk, Joseph W.; Brow, Richard K.; Zhang, San-Hong

    2016-02-01

    Corrosion-resistant Ni3(Si,Nb) alloys are promising materials of construction for hydrogen-production systems based on the sulfur-iodine thermochemical cycle. In this work, the corrosion rates of three different Ni3(Si,Nb) alloys were measured in boiling 70 wt.% sulfuric acid and a three-stage corrosion mechanism was identified, based on the composition and morphology of surface scale that developed. The α(Ni) + β(Ni3Si) eutectic constituent of the alloy microstructure was selectively attacked by acid and, when present, is detrimental to corrosion resistance. The G-phase (Ni16Si17Nb6) is more passive than the β-matrix and seems to contribute to a lower steady-state corrosion rate.

  11. The Results of Feasibility Study of Co-generation NPP With Innovative VK-300 Simplified Boiling Water Reactor

    SciTech Connect

    Kuznetsov, Yury N.

    2006-07-01

    The co-generation nuclear power plant (CNPP) producing electricity and district heating heat is planned to be constructed in Archangelsk Region of Russia. Following the 'Letter of Intent' signed by Governor of Archangelsk region and by Minister of the Russian Federation for atomic energy the feasibility study of the Project has been done. The NPP will be based on the four co-generation nuclear power units with the Russian VK-300 SBWR. The innovative passive VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype VK-50 reactor in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR-1000 (Siemens). The CNPP's total power is planned to be 1000 MW(e) and district-heating heat production capacity 1600 Gcal/h. A detailed description of the results of the feasibility study is presented in the report. The results of the feasibility study have shown that the Archangelsk CGNP is feasible in terms of engineering, economics and production. (authors)

  12. 78 FR 58575 - Review of Experiments for Research Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-24

    ... COMMISSION Review of Experiments for Research Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Guide (RG) 2.4, ``Review of Experiments for Research Reactors.'' The guide is being withdrawn because... Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because its guidance no longer...

  13. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    SciTech Connect

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  14. Assessment of light water reactor accident management programs and experience

    SciTech Connect

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  15. Dryout and Rewetting in the Pool Boiling Experiment Flown on STS-72 (PBE-2 B) and STS-77 (PBE-2 A)

    NASA Technical Reports Server (NTRS)

    Merte, Herman, Jr.; Lee, Ho Sung; Keller, Robert B.

    1998-01-01

    Experiments were conducted in the microgravity of space in which a pool of liquid (R-113), initially at a precisely defined pressure and temperature, is subjected to a step imposed heat flux from a semi-transparent thin-film heater forming part of one wall of the container such that boiling is initiated and maintained for a defined period of time at a constant pressure level. A total of nine tests were conducted at three levels of heat flux and three levels of subcooling in each of the two space experiments in a GAS canister on the STS-77, -72, respectively. Three (3) modes of propagation of boiling across the heater surface and subsequent vapor bubble growths were observed, in addition to the two (2) modes observed in the previous microgravity pool boiling space flights on STS-47, -57, and -60. Of particular interest were the extremely dynamic or "explosive" growths, which were determined to be the consequence of the large increase in the liquid-vapor interface area associated with the appearance of a corrugated or rough interface. Predictions of circumstances for its onset have been carried out. Assumptions were necessary regarding the character of disturbances necessary for the instabilities to grow. Also, a new vapor bubble phenomena was observed in which small vapor bubbles migrated toward a larger bubble, eventually coalescing with this larger bubble. The heat transfer was enhanced approximately 30% as a result of these migrating bubbles, which is believed to be a vapor bubble manifestation of Marangoni convection and/or molecular momentum effects, sometimes referred to as vapor recoil. The circumstances of heat flux and liquid subcooling necessary to produce heater surface dryout for an initially stagnant liquid subjected to an imposed heat flux have been more closely identified.

  16. BWRSAR (Boiling Water Reactor Severe Accident Response) calculations of reactor vessel debris pours for Peach Bottom short-term station blackout

    SciTech Connect

    Hodge, S.A.; Ott, L.J.

    1988-01-01

    This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident Response (BWRSAR) code and are based upon consideration of the Peach Bottom Atomic Power Station. The modeling strategies employed within BWRSAR for debris relocation within the reactor vessel are briefly discussed and the calculated events of the accident sequence, including details of the calculated debris pours, are presented. 4 refs., 13 figs., 3 tabs.

  17. Visualization study on pool boiling heat transfer

    NASA Astrophysics Data System (ADS)

    Kamei, Shuya; Hirata, Masaru

    1991-04-01

    The visualized boiling phenomena were observed by means of high speed photographic shadowgraphy using a rotating prism camera (nac HIGH SPEED CAMERA model-16HD) with the speed of about 3500 frames per second. The photographs show that pool boiling heat transfer phenomena are varied for the boiling curve based on the experiments. Experiments have been carried out to investigate pool boiling heat transfer phenomena on a horizontal thin filament in subcooled and saturated distilled water. The experiments were performed for atmospheric pressure,for filament diameters of about 0.3 mm, for region of natural convection to film boiling. The color-film made by high speed movie camera are converted to high speed color video-tape. It is convenient to edit and show the tape for visualization with teaching the students. The high speed color video showed that the successive motion and shape of bubbles during their process of detachment varied with increasing heat flux on the heated surface of a filament. From these results, it was confirmed that the high speed phenomena of boiling by the slow motion video pictures could be estimated clearly.

  18. A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

    ERIC Educational Resources Information Center

    Derevjanik, Mario; Badri, Solmaz; Barat, Robert

    2011-01-01

    This experiment and analysis offer an economic yet challenging semi-batch reactor experience. Household bleach is pumped at a controlled rate into a batch reactor containing pharmaceutical hydrogen peroxide solution. Batch temperature, product molecular oxygen, and the overall change in solution conductivity are metered. The reactor simulation…

  19. IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ion chamber beneath reactor on dolly. Photographer: Page Comiskey. Date: August 11, 1965. INEEL negative no. 65-4039 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  20. The radon monitoring system in Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Chu, M. C.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Leung, J. K. C.; Leung, K. Y.; Lin, Y. C.; Luk, K. B.; Pun, C. S. J.

    2016-02-01

    We developed a highly sensitive, reliable and portable automatic system (H3) to monitor the radon concentration of the underground experimental halls of the Daya Bay Reactor Neutrino Experiment. H3 is able to measure radon concentration with a statistical error less than 10% in a 1-h measurement of dehumidified air (R.H. 5% at 25 °C) with radon concentration as low as 50 Bq/m3. This is achieved by using a large radon progeny collection chamber, semiconductor α-particle detector with high energy resolution, improved electronics and software. The integrated radon monitoring system is highly customizable to operate in different run modes at scheduled times and can be controlled remotely to sample radon in ambient air or in water from the water pools where the antineutrino detectors are being housed. The radon monitoring system has been running in the three experimental halls of the Daya Bay Reactor Neutrino Experiment since November 2013.

  1. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces.

  2. Film boiling on spheres in single- and two-phase flows.

    SciTech Connect

    Liu, C.; Theofanous, T. G.

    2000-08-29

    Film boiling on spheres in single- and two-phase flows was studied experimentally and theoretically with an emphasis on establishing the film boiling heat transfer closure law, which is useful in the analysis of nuclear reactor core melt accidents. Systematic experimentation of film boiling on spheres in single-phase water flows was carried out to investigate the effects of liquid subcooling (from 0 to 40 C), liquid velocity (from 0 to 2 m/s), sphere superheat (from 200 to 900 C), sphere diameter (from 6 to 19 mm), and sphere material (stainless steel and brass) on film boiling heat transfer. Based on the experimental data a general film boiling heat transfer correlation is developed. Utilizing a two-phase laminar boundary-layer model for the unseparated front film region and a turbulent eddy model for the separated rear region, a theoretical model was developed to predict the film boiling heat transfer in all single-phase regimes. The film boiling from a sphere in two-phase flows was investigated both in upward two-phase flows (with void fraction from 0.2 to 0.65, water velocity from 0.6 to 3.2 m/s, and steam velocity from 3.0 to 9.0 m/s) and in downward two-phase flows (with void fraction from 0.7 to 0.95, water velocity from 1.9 to 6.5 m/s, and steam velocity from 1.1 to 9.0 m/s). The saturated single-phase heat transfer correlation was found to be applicable to the two-phase film boiling data by making use of the actual water velocity (water phase velocity), and an adjustment factor of (1 - {alpha}){sup 1/4} (with a being the void fraction) for downward flow case only. Slight adjustments of the Reynolds number exponents in the correlation provided an even better interpretation of the two-phase data. Preliminary experiments were also conducted to address the influences of multi-sphere structure on the film boiling heat transfer in single- and two-phase flows.

  3. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    NASA Astrophysics Data System (ADS)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  4. The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

    SciTech Connect

    Mitenkova, E. F.; Novikov, N. V.; Blokhin, A. I.

    2012-07-01

    The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

  5. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  6. ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3. 06. 6B - transient film boiling in upflow. [PWR

    SciTech Connect

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  7. Some scoping experiments for a space reactor

    SciTech Connect

    Alexander, C.A.; Ogden, J.S.

    1983-07-07

    Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failure exists with UO/sub 2/ and a lithium heat pipe. Changing the composition of the metal of the heat pipe would have only a second order effect on the kinetics of the failure mechanism. Uranium carbide and nitride were considered as potential fuels which are nonreactive in a lithium environment. At high temperatures the nitride would be favored because of its better compatibility with potential cladding materials. Compositions of UN with small additions of YN appear to offer very attractive properties for a compact high temperature high power density reactor.

  8. RENO-CC: A new system to fuel lattice design in boiling water reactors using neural networks

    SciTech Connect

    Ortiz, J. J.; Perusquia, R.; Hernandez, J. L.

    2006-07-01

    We show a new system to optimize fuel lattices in BWRs named RENO-CC. The system employs a multi state recurrent neural network (MSRNN) for optimizing a fuel lattice pin-by-pin {sup 235}U enrichment distribution. Local Power Peaking Factor (LPPF) and k-infinite (k{sub {infinity}}) are involved in the MSRNN energy function. Both parameters are calculated by the 2D HELIOS transport code for lattice burn-up. Through the iterative process the MSRNN decreases PPF value while k{sub {infinity}}, is kept in a rank of values, at the beginning of lattice life (BOL). The iterative process ends after 20 iterations. If PPF is not lower than limit, RENO-CC applies a fuzzy logic rule in order to recommend if the fuel lattice has an acceptable LPPF value and it might eventually be used in a fuel load. When a fuel lattice is obtained it can be used into a fuel assembly. And eventually, this fuel assembly would be used in the process of fuel load and control rod patterns optimization. So, a 3D core reactor calculation must decide if such a lattice design can fulfill the operation conditions into the reactor core. Preliminary results are shown in this paper. (authors)

  9. Transient pool boiling in microgravity

    NASA Technical Reports Server (NTRS)

    Ervin, J. S.; Merte, H., Jr.; Keller, R. B.; Kirk, K.

    1992-01-01

    Transient nucleate pool boiling experiments using R113 are conducted for short times in microgravity and in earth gravity with different heater surface orientations and subcoolings. The heating surface is a transparent gold film sputtered on a quartz substrate, which simultaneously provides surface temperature measurements and permits viewing of the boiling process from beneath. For the microgravity experiments, which have uniform initial temperatures and no fluid motion, the temperature distribution in the R 113 at the moment of boiling inception is known. High speed cameras with views both across and through the heating surface record the boiling spread across the heater surface, which is classified into six distinct categories.

  10. Advances in Computational Fluid Dynamics Modeling of Two Phase Flow in a Boiling Water Reactor Fuel Assembly

    SciTech Connect

    Tentner, Adrian; Lo, Simon; Ioilev, Andrey; Melnikov, Vladimir; Samigulin, Maskhud; Ustinenko, Vasily; Kozlov, Valentin

    2006-07-01

    A new code, CFD-BWR, is being developed for the simulation of two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. CFD-BWR is a specialized module built on the foundation of the commercial CFD code STAR-CD which provides general two-phase flow modeling capabilities. New models describing the inter-phase mass, momentum, and energy transfer phenomena specific for BWRs have been developed and implemented in the CFD-BWR module. A set of experiments focused on two-phase flow and phase-change phenomena has been identified for the validation of the CFD-BWR code and results of two experiment analyses focused on the radial void distribution are presented. The close agreement between the computed results, the measured data and the correlation results provides confidence in the accuracy of the models. (authors)

  11. Investigation of the physical and numerical foundations of two-fluid representation of sodium boiling with applications to LMFBR experiments

    SciTech Connect

    No, H.C.; Kazimi, M.S.

    1983-03-01

    This work involves the development of physical models for the constitutive relations of a two-fluid, three-dimensional sodium boiling code, THERMIT-6S. The code is equipped with a fluid conduction model, a fuel pin model, and a subassembly wall model suitable for stimulating LMFBR transient events. Mathematically rigorous derivations of time-volume averaged conservation equations are used to establish the differential equations of THERMIT-6S. These equations are then discretized in a manner identical to the original THERMIT code. A virtual mass term is incorporated in THERMIT-6S to solve the ill-posed problem. Based on a simplified flow regime, namely cocurrent annular flow, constitutive relations for two-phase flow of sodium are derived. The wall heat transfer coefficient is based on momentum-heat transfer analogy and a logarithmic law for liquid film velocity distribution. A broad literature review is given for two-phase friction factors. It is concluded that entrainment can account for some of the discrepancies in the literature. Mass and energy exchanges are modelled by generalization of the turbulent flux concept. Interfacial drag coefficients are derived for annular flows with entrainment. Code assessment is performed by simulating three experiments for low flow-high power accidents and one experiment for low flow/low power accidents in the LMFBR. While the numerical results for pre-dryout are in good agreement with the data, those for post-dryout reveal the need for improvement of the physical models. The benefits of two-dimensional non-equilibrium representation of sodium boiling are studied.

  12. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    SciTech Connect

    Bierschbach, M.C.

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  13. Boiling Temperature and Reversed Deliquescence Relative Humidity Measurements for Mineral Assemblages in the NaCl + NaNO3 + KNO3 + Ca(NO3)2 + H2O System

    SciTech Connect

    Rard, J A; Staggs, K J; Day, S D; Carroll, S A

    2005-12-01

    Boiling temperature measurements have been made at ambient pressure for saturated ternary solutions of NaCl + KNO{sub 3} + H{sub 2}O, NaNO{sub 3} + KNO{sub 3} + H{sub 2}O, and NaCl + Ca(NO{sub 3}){sub 2} + H{sub 2}O over the full composition range, along with those of the single salt systems. Boiling temperatures were also measured for the four component NaCl + NaNO{sub 3} + KNO{sub 3} + H{sub 2}O and five component NaCl + NaNO{sub 3} + KNO{sub 3} + Ca(NO{sub 3}){sub 2} + H{sub 2}O mixtures, where the solute mole fraction of Ca(NO{sub 3}){sub 2}, x(Ca(NO{sub 3}){sub 2}), was varied between 0 and 0.25. The maximum boiling temperature found for the NaCl + KNO{sub 3} + H{sub 2}O system is {approx} 134.9 C; for the NaNO{sub 3} + KNO{sub 3} + H{sub 2}O system is {approx} 165.1 C at x(NaNO{sub 3}) {approx} 0.46 and x(KNO{sub 3}) {approx} 0.54; and for the NaCl + Ca(NO{sub 3}){sub 2} + H{sub 2}O system is 164.7 {+-} 0.6 C at x(NaCl) {approx} 0.25 and x(Ca(NO{sub 3}){sub 2}) {approx} 0.75. The NaCl + NaNO{sub 3} + KNO{sub 3} + Ca(NO{sub 3}){sub 2} + H{sub 2}O system forms molten salts below their maximum boiling temperatures, and the temperatures corresponding to the cessation of boiling (dry out temperatures) of these liquid mixtures were determined. These dry out temperatures range from {approx} 300 C when x(Ca(NO{sub 3}){sub 2}) = 0 to {ge} 400 C when x(Ca(NO{sub 3}){sub 2}) = 0.20 and 0.25. Mutual deliquescence/efflorescence relative humidity (MDRH/MERH) measurements were also made for the NaNO{sub 3} + KNO{sub 3} and NaCl + NaNO{sub 3} + KNO{sub 3} salt mixture from 120 to 180 C at ambient pressure. The NaNO{sub 3} and NaCl + NaNO{sub 3} + KNO{sub 3} salt mixture has a MDRH of 26.4% at 120 C and 20.0% at 150 C. This salt mixture also absorbs water at 180 C, which is higher than expected from the boiling temperature experiments. The NaCl + NaNO{sub 3} + KNO{sub 3} salt mixture was found to have a MDRH of 25.9% at 120 C and 10.5% at 180 C. The investigated mixture

  14. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    Energy Science and Technology Software Center (ESTSC)

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized watermore » reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.« less

  15. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    SciTech Connect

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized water reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.

  16. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    SciTech Connect

    Not Available

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  17. On the shape of stress corrosion cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant piping at 288 °C

    NASA Astrophysics Data System (ADS)

    Lee, Sang-Kwon; Kramer, Daniel; Macdonald, Digby D.

    2014-11-01

    Evolution of the shape of surface cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant circuit piping at the reactor operating temperature of 288 °C is explored as a function of various environmental variables, such as electrochemical potential (ECP), solution conductivity, flow velocity, and multiplier for the oxygen reduction reaction (ORR) standard exchange current density (SECD), using the coupled environment fracture model (CEFM). For this work, the CEFM was upgraded by incorporating Shoji's model for calculating the crack tip strain rate and more advanced expressions were used for estimating the stress intensity factor for semi-elliptical surface cracks. This revised CEFM accurately predicts the dependence of the crack growth rate on stress intensity factor and offers an alternative explanation for the development of semi-elliptical cracks than that provided by fracture mechanics alone. The evolution of surface crack semi-elliptical shape depends strongly upon various environmental variables identified above, and the CEFM predicts that the minor axis of the ellipse should be oriented perpendicular to the surface, in agreement with observation. The development of the observed semi-elliptical cracks with the minor axis perpendicular to the surface is therefore attributed to the dependence of the crack growth rate on the electrochemical crack length.

  18. Consort 3 flight experiments

    NASA Technical Reports Server (NTRS)

    Wessling, Francis C.; Maybee, George W.

    1991-01-01

    The third sounding rocket payload in the Consort program was launched from the White Sands Missile Range on May 16, 1990. It carried 12 experiments designed to investigate materials processes in low gravity. All of the experiments were reflights from the Consort 2 payload that was launched on November 1989 but failed to achieve microgravity because of a malfunction in the launch vehicle. Four national Centers for the Commercial Development of Space participated in the mission. The payload included five experiments and two accelerometer systems that flew on Consort 1 and seven new experiments, designed and developed since Consort 1. Experiments from Consort 1 incorporated hardware modifications and changes in experimental parameters based on mission results. The new experiments covered a variety of polymeric and biological investigations. A new power distribution and control system designed to provide discrete, computer-supervised, experiment power monitoring and control was flight qualified on Consort 3. Consort 3 featured very late access (3-5 h before launch vs 27-30 h for Consort 1) for installation of sensitive biological specimens. The integrated payload and mission sequence of events are described. Changes in the Consort 1 experiments are defined and the objectives, methods, and expectations for new experiments are discussed.

  19. Recent reactor testing and experience with gamma thermometers

    SciTech Connect

    Waring, J.P.; Smith, R.D.

    1983-02-01

    Recent experience with gamma thermometers for light water reactors has primarily been in the Framatome reactors operated by Electricite de France. Other recent testing has taken place at Oak Ridge National Laboratory and the Otto Hahn ship reactor. Earlier experience with gamma thermometers was in heavy water reactors at Savannah River and Halden. This paper presents recent data from the light water reactor (LWR) programs. The principles of design and operation of the Radcal gamma thermometer were presented in ''Gamma Thermometer Developments for Light Water Reactors'', Leyse and Smith/sup 1/. Observations from LWRs confirm the earlier experience from heavy water reactors that the gamma thermometer units give signals which are proportional to the power of surrounding fuel rods and virtually independent of exposure, surrounding poison and other conditions which affect signals of neutron sensitive devices. After 200 sensor-years in EdF reactors, there has been no change in the sensitivity of the devices. Nonetheless, the Radcal units can be recalibrated in-reactor by the introduction of electrical heating via a heater cable imbedded in the device. Algorithms and signal processing software have been developed to interpret and display the gamma thermometer signals. The results of this processing are illustrated here.

  20. Review on θ13 Measurements from Reactor Experiments

    NASA Astrophysics Data System (ADS)

    Kim, Soo-Bong

    2015-03-01

    Definitive measurements of the smallest neutrino mixing angle θ13 were made by Daya Bay, Double Chooz and RENO experiments in 2012, based on the disappearance of electron antineutrinos emitted from reactors. The new generation reactor experiments have significantly improved a sensitivity for θ13 down to the sin2(2θ13) ~ 0.01 level. A rather large value of θ13 has opened a new window to find the CP violation phase and to determine the neutrino mass hierarchy. The θ13 measurements by the three reactor experiments are presented with their future expected sensitivities.

  1. Characteristics of Transient Boiling Heat Transfer

    SciTech Connect

    Liu, Wei; Monde, Masanori; Mitsutake, Y.

    2002-07-01

    In this paper, one dimensional inverse heat conduction solution is used for a measurement of pool boiling curve. The experiments are performed under atmospheric pressure for copper, brass, carbon steel and gold. Boiling curves, including unsteady transition boiling region, are found can be traced fairly well from a simple experiment system by solving inverse heat conduction solution. Boiling curves for steady heating and transient heating, for heating process and cooling process are compared. Surface behavior around CHF point, transition boiling and film-boiling regions are observed by using a high-speed camera. The results show the practicability of the inverse heat conduction solution in tracing boiling curve and thereby supply us a new way in boiling heat transfer research. (authors)

  2. Comparison of Computational Results with a Low-g, Nitrogen Slosh and Boiling Experiment

    NASA Technical Reports Server (NTRS)

    Stewart, Mark; Moder, Jeff

    2015-01-01

    The proposed paper will compare a fluid/thermal simulation, in FLUENT, with a low-g, nitrogen slosh experiment. The French Space Agency, CNES, performed cryogenic nitrogen experiments in several zero gravity aircraft campaigns. The computational results have been compared with high-speed photographic data, pressure data, and temperature data from sensors on the axis of the cylindrically shaped tank. The comparison between these experimental and computational results is generally favorable: the initial temperature stratification is in good agreement, and the two-phase fluid motion is qualitatively captured.

  3. An Improved Design of a Simple Tubular Reactor Experiment.

    ERIC Educational Resources Information Center

    Asfour, Abdul-Fattah A.

    1985-01-01

    Background information, procedures used, and typical results obtained are provided for an experiment which: (1) examines the effect of residence time on conversion in a tubular flow reactor; and (2) compares the experimental conversions with those obtained from plug-flow and laminar-flow reactor models. (JN)

  4. 139. ARAIII Index of drwaings of gascooled reactor experiment buildings. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    139. ARA-III Index of drwaings of gas-cooled reactor experiment buildings. Aerojet-general 880-area/GCRE-100. Date: February 1958. Ineel index code no. 063-9999-80-013-102505. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  5. Experimental parameters for a reactor antineutrino experiment at very short baselines

    NASA Astrophysics Data System (ADS)

    Heeger, K. M.; Tobin, M. N.; Littlejohn, B. R.; Mumm, H. P.

    2013-04-01

    Reactor antineutrinos are used to study neutrino oscillation, search for signatures of nonstandard neutrino interactions, and monitor reactor operation for safeguard applications. The flux and energy spectrum of reactor antineutrinos can be predicted from the decays of the nuclear fission products. A comparison of recent reactor calculations with past measurements at baselines of 10-100 m suggests a 7.2% deficit. Precision measurements of reactor antineutrinos at very short baselines O(1-10m) can be used to probe this anomaly and search for possible oscillations into sterile neutrino species. This paper studies the experimental requirements for a new reactor antineutrino measurement at very short baselines and calculates the sensitivity of various scenarios. We conclude that, given proper site optimization, detector design, and background reduction, an experiment at a typical research reactor can provide 5σ discovery potential for the favored oscillation parameter space with 3 years of detector live time.

  6. Physical Science Informatics: Providing Open Science Access to Microheater Array Boiling Experiment Data

    NASA Technical Reports Server (NTRS)

    McQuillen, John; Green, Robert D.; Henrie, Ben; Miller, Teresa; Chiaramonte, Fran

    2014-01-01

    The Physical Science Informatics (PSI) system is the next step in this an effort to make NASA sponsored flight data available to the scientific and engineering community, along with the general public. The experimental data, from six overall disciplines, Combustion Science, Fluid Physics, Complex Fluids, Fundamental Physics, and Materials Science, will present some unique challenges. Besides data in textual or numerical format, large portions of both the raw and analyzed data for many of these experiments are digital images and video, requiring large data storage requirements. In addition, the accessible data will include experiment design and engineering data (including applicable drawings), any analytical or numerical models, publications, reports, and patents, and any commercial products developed as a result of the research. This objective of paper includes the following: Present the preliminary layout (Figure 2) of MABE data within the PSI database. Obtain feedback on the layout. Present the procedure to obtain access to this database.

  7. Reduced gravity boiling and condensing experiments simulated with the COBRA/TRAC computer code

    SciTech Connect

    Cuta, J.M.; Krotiuk, W.J.

    1988-02-01

    It is being recognized that there does not currently exist an adequate understanding of flow and heat transfer behavior in reduced- and zero-gravity. There is not a sufficient experimental fluid-thermal data base to develop design correlations for two-phase pressure losses, heat transfer coefficients, and critical heat flux limits in systems proposed for advanced power sources, propulsion, and other thermal management systems in space. Pacific Northwest Laboratory (PNL), is the lead laboratory for thermal hydraulics in the Department of Energy's Multimegawatt Space Power Program, and has the responsibility of developing microgravity thermal-hydraulic analysis capabilities for application to space nuclear power systems. In support of this program, PNL has performed a series of reduced-gravity two-phase flow experiments in the NASA KC-135 aircraft. The objective of the experiment was to supply basic thermal-hydraulic information that could be used in development of analytical design tools. 6 refs., 23 figs., 4 tabs.

  8. Reduced gravity boiling and condensing experiments simulated with the COBRA/TRAC computer code

    NASA Technical Reports Server (NTRS)

    Cuta, Judith M.; Krotiuk, William

    1988-01-01

    A series of reduced-gravity two-phase flow experiments has been conducted with a boiler/condenser apparatus in the NASA KC-135 aircraft in order to obtain basic thermal-hydraulic data applicable to analytical design tools. Several test points from the KC-135 tests were selected for simulation by means of the COBRA/TRAC two-fluid, three-field thermal-hydraulic computer code; the points were chosen for a 25-90 percent void-fraction range. The possible causes for the lack of agreement noted between simulations and experiments are explored, with attention to the physical characteristics of two-phase flow in one-G and near-zero-G conditions.

  9. ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3. 07. 9 - steady-state film boiling in upflow

    SciTech Connect

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  10. System Requirements Document for the Molten Salt Reactor Experiment

    SciTech Connect

    Aigner, R.D.

    2000-04-01

    The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

  11. Results on θ13 Neutrino Oscillations from Reactor Experiments

    NASA Astrophysics Data System (ADS)

    Kim, Soo-Bong

    2014-03-01

    Definitive measurements of the smallest neutrino mixing angle θ13 were made by Daya Bay, Double Chooz and RENO in 2012, based on the disappearance of electron antineutrinos emitted from reactors. The new generation reactor experiments have significantly improved a sensitivity for θ13 down to the sin2(2θ13)~0.01 level using two identical detectors of 10 ~ 40 tons at near (300 ~ 400 m) and far (1 ~ 2 km) locations. The θ13 measurements by the three reactor experiments are presented with their future expected sensitivities.

  12. An overview of the Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    Cao, Jun; Luk, Kam-Biu

    2016-07-01

    The Daya Bay Reactor Neutrino Experiment discovered an unexpectedly large neutrino oscillation related to the mixing angle θ13 in 2012. This finding paved the way to the next generation of neutrino oscillation experiments. In this article, we review the history, featured design, and scientific results of Daya Bay. Prospects of the experiment are also described.

  13. Summary Results of the Neptun Boil-Off Experiments to Investigate the Accuracy and Cooling Influence of LOFT Cladding-Surface Thermocouples (System 00)

    SciTech Connect

    E. L. Tolman S. N. Aksan

    1981-10-01

    Nine boil-off experiments were conducted in the Swiss NEPTUN Facility primarily to obtain experimental data for assessing the perturbation effects of LOFT thermocouples during simulated small-break core uncovery conditions. The data will also be useful in assessing computer model capability to predict thermal hydraulic response data for this type of experiment. System parameters that were varied for these experiments included heater rod power, system pressure, and initial coolant subcooling. The experiments showed that the LOFT thermocouples do not cause a significant cooling influence in the rods to which they are attached. Furthermore, the accuracy of the LOFT thermocouples is within 20 K at the peak cladding temperature zone.

  14. Plasma Reactor Modeling and Validation Experiments

    NASA Technical Reports Server (NTRS)

    Meyyappan, M.; Bose, D.; Hash, D.; Hwang, H.; Cruden, B.; Sharma, S. P.; Rao, M. V. V. S.; Arnold, Jim (Technical Monitor)

    2001-01-01

    Plasma processing is a key processing stop in integrated circuit manufacturing. Low pressure, high density plum reactors are widely used for etching and deposition. Inductively coupled plasma (ICP) source has become popular recently in many processing applications. In order to accelerate equipment and process design, an understanding of the physics and chemistry, particularly, plasma power coupling, plasma and processing uniformity and mechanism is important. This understanding is facilitated by comprehensive modeling and simulation as well as plasma diagnostics to provide the necessary data for model validation which are addressed in this presentation. We have developed a complete code for simulating an ICP reactor and the model consists of transport of electrons, ions, and neutrals, Poisson's equation, and Maxwell's equation along with gas flow and energy equations. Results will be presented for chlorine and fluorocarbon plasmas and compared with data from Langmuir probe, mass spectrometry and FTIR.

  15. Experience in the reprocessing of mixed-oxide fuels at PNC (Power Reactor and Nuclear Fuel Development Corporation)

    SciTech Connect

    Komatsu, Hisato; Onishi, Moichi; Araya, Sadao; Fukushima, Misao

    1989-01-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan has experience in reprocessing mixed-oxide (MOX) fuels for the advanced thermal reactor (ATR) Fugen at the Tokai Reprocessing Plant (TRP) and for fast breeder reactors (FBRs) at the Chemical Processing Facility (CPF). The TRP was originally designed and constructed as the first reprocessing plant for light water reactor fuels in Japan. It has processed {approximately}400 t of spent fuels since 1977. To utilize recovered plutonium, PNC has developed the prototype ATR Fugen. This reactor has been operated using MOX fuel since 1978. In parallel, utilities are promoting a plutonium thermal project. Several MOX assemblies have already been loaded in a boiling water and a pressurized water reactor. To facilitate the operation of Fugen and promote research and development for the reprocessing of MOX fuels in Japan, PNC obtained a license for reprocessing fuels for Fugen at TRP in 1985. PNC has designed and constructed the CPF at Tokai Works to conduct basic research on the reprocessing of FBR fuels. The Recycle Equipment Test Facility, an engineering scale hot facility, is now being designed for further R and D in this field. It will start hot operation in the mid-1990s.

  16. A New Method to Determine the Neutron Sensitivity of a Micro Fission Chamber for a Boiling Water Reactor

    SciTech Connect

    Kimura, Atsushi; Fujita, Kaoru; Nakazawa, Masaharu; Fujita, Shinya; Seki, Eiji; Tanaka, Yutaka; Kono, Shigehiro

    2001-11-15

    A new method to determine neutron sensitivity of uranium-type local power range monitors (LPRMs) has been developed. In this method, neutron sensitivity of LPRMs is obtained from alpha-ray current of {sup 234}U. The uncertainty of the calibration for neutron sensitivity using alpha-ray current was 2.1%, and the correlation coefficient of these neutron sensitivities was 0.868 (sample: 50 cases). Using this method, the neutron sensitivity can be obtained without irradiation tests of LPRMs in reactors.

  17. Double Chooz and a history of reactor θ13 experiments

    NASA Astrophysics Data System (ADS)

    Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

    2016-07-01

    This is a contribution paper from the Double Chooz (DC) experiment to the special issue of Nuclear Physics B on the topics of neutrino oscillations, celebrating the recent Nobel prize to Profs. T. Kajita and A.B. McDonald. DC is a reactor neutrino experiment which measures the last neutrino mixing angle θ13. The DC group presented an indication of disappearance of the reactor neutrinos at a baseline of ∼1 km for the first time in 2011 and is improving the measurement of θ13. DC is a pioneering experiment of this research field. In accordance with the nature of this special issue, physics and history of the reactor-θ13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.

  18. Experiment on nucleate pool boiling in microgravity by using transparent heating surface - Analysis of surface heat transfer coefficients

    NASA Astrophysics Data System (ADS)

    Kubota, C.; Kawanami, O.; Asada, Y.; Wada, Y.; Nagayasu, T.; Shinmoto, Y.; Ohta, H.; Kabov, O.; Queeckers, P.; Chikov, S.; Straub, J.

    2011-12-01

    Investigation of mechanisms in nucleate boiling under microgravity conditions is essential for the development of the cooling systems handling a large amount of waste heat. A transparent heating surface with multiple arrays of 88 thin film temperature sensors and mini-heaters was developed for the clarification of boiling heat transfer mechanisms in microgravity. To investigate gravity effects on the microlayer behaviors and corresponding local heat transfer coefficients, images of liquid-vapor behaviors underneath attached bubbles and local heat transfer data were simultaneously obtained in microgravity pool boiling. The present paper reports the analysis of the data measured during the ESA parabolic flight campaign. It was found that the liquid-vapor behaviors were strongly affected by the direction and the level of residual gravity. Various patterns of liquid-vapor behaviours and corresponding enhancement or deterioration of the heat transfer are observed.

  19. The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

    SciTech Connect

    John D. Bess; Michael A. Pope; Harold F. McFarlane

    2012-11-01

    The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactor (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.

  20. PROSPECT: A Precision Reactor Oscillation and Spectrum Experiment

    NASA Astrophysics Data System (ADS)

    Littlejohn, Bryce

    2014-03-01

    Antineutrino detectors operated close to a compact reactor core can provide excellent sensitivity to short-baseline oscillation effects by precisely measuring any relative distortion of the ν̲e spectrum as a function of both energy and baseline. Such a measurement can be performed in the United States at several highly-enriched uranium fueled research reactors using near-surface segmented scintillator detectors. This talk will describe the preliminary conceptual design and oscillation physics potential of the PROSPECT experiment, a U.S.-based, multi-phase, 2-detector experiment with reactor-detector baselines of 4-20 meters capable of excluding a majority of the suggested sterile neutrino oscillation parameter space at high confidence level. Additional goals, such as precise measurement of the ν̲e spectrum from a highly-enriched uranium core, as well as development of detection techniques and technology for reactor monitoring applications, will be discussed.

  1. Cosmic muon background and reactor neutrino detectors: the Angra experiment

    NASA Astrophysics Data System (ADS)

    Casimiro, E.; Anjos, J. C.

    2008-06-01

    We discuss on the importance of appropriately taking into account the cosmic background in the design of reactor neutrino detectors. In particular, as a practical study case, we describe the Angra Project, a new reactor neutrino oscillation experiment proposed to be built in the coming years at the Brazilian nuclear power complex, located near the Angra dos Reis city. The main goal of the experiment is to measure with high precision θ13, the last unknown of the three neutrino mixing angles. The experiment will in addition explore the possibility of using neutrino detectors for purposes of safeguards and non-proliferation of nuclear weapons.

  2. Results from Boiling Temperature Measurements for Saturated Solutions in the Systems NaCl + Ca(NO3)2 + H2O, NaNO3 + KNO3 + H2O, and NaCl + KNO3 + H2O, and Dry Out Temperatures for NaCl + NaNO3 + KNO3 + Ca(NO3)2 + H2O

    SciTech Connect

    Rard, J A

    2005-11-29

    Boiling temperature measurements have been made for saturated ternary solutions of NaCl + KNO{sub 3} + H{sub 2}O and NaNO{sub 3} + KNO{sub 3} + H{sub 2}O at three selected salt ratios and for NaCl + Ca(NO{sub 3}){sub 2} + H{sub 2}O over the full composition range. The maximum boiling temperature found for the NaCl + Ca(NO{sub 3}){sub 2} + H{sub 2}O system is 164.7 {+-} 0.6 C, and the composition is estimated to occur at x(Ca(NO{sub 3}){sub 2}) {approx} 0.25. Experiments were also performed for the five component NaCl + NaNO{sub 3} + KNO{sub 3} + Ca(NO{sub 3}){sub 2} + H{sub 2}O mixtures with the molar ratio of NaCl:NaNO{sub 3}:KNO{sub 3} held essentially constant at 1:0.9780:1.1468 as the solute mole fraction of Ca(NO{sub 3}){sub 2}, x(Ca(NO{sub 3}){sub 2}), was varied between 0 and 0.25. The NaCl + NaNO{sub 3} + KNO{sub 3} + Ca(NO{sub 3}){sub 2} + H{sub 2}O system forms low melting mixtures and thus boiling temperatures for saturated were not determined. Instead, the temperatures corresponding to the cessation of boiling (i.e., dry out temperatures) of these liquid mixtures were determined. These dry out temperatures range from {approx} 300 C when x(Ca(NO{sub 3}){sub 2}) = 0 to {ge} 400 C when x(Ca(NO{sub 3}){sub 2}) = 0.20 and 0.25. The investigated mixture compositions correspond to some of the major mineral assemblages that are predicted to control the deliquescence relative humidity of salts formed by leaching dust samples from the proposed nuclear repository at Yucca Mountain, Nevada.

  3. Liquid metal boiling inception

    NASA Technical Reports Server (NTRS)

    Sabin, C. M.; Poppendiek, H. F.; Mouritzen, G.; Meckel, P. T.; Cloakey, J. E.

    1972-01-01

    An experimental study of the inception of boiling in potassium in forced convection is reported. The boiler consisted of a 0.19-inch inside diameter, niobium-1% zirconium boiler tube approximately six feet long. Heating was accomplished by direct electrical tube wall conduction. Experiments were performed with both all-liquid fill and two-phase fill startup sequences and with a range of flow rates, saturation temperatures, inert gas levels, and fill liquid temperatures. Superheat of the liquid above the equilibrium saturation temperature was observed in all the experiments. Incipient boiling liquid superheat ranged from a few degrees to several hundred. Comparisons of these data with other data and with several analytical treatments are presented.

  4. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  5. Use of an influence diagram and fuzzy probability for evaluating accident management in a boiling water reactor

    SciTech Connect

    Yu, D.; Kastenberg, W.E.; Okrent, D. . Mechanical, Aerospace, and Nuclear Engineering Dept.)

    1994-06-01

    A new approach is presented for evaluating the uncertainties inherent in severe accident management strategies. At first, this analysis considers accident management as a decision problem (i.e., applying a strategy compared with do nothing) and uses an influence diagram. To evaluate imprecise node probabilities in the influence diagram, the analysis introduces the concept of a fuzzy probability. When fuzzy logic is applied, fuzzy probabilities are easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach, which uses point-estimate values, but also additional information regarding the impact of using imprecise input data. As an illustrative example, the proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence at the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy is beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of containment failure for both liner melt-through and late overpressurization. Even though uncertainty exists in the results, flooding is preferred to do nothing when evaluated in terms of two risk measures: early and late fatalities.

  6. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    SciTech Connect

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. ); Wagner, K.C. )

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  7. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  8. Evaluation of performance of select fusion experiments and projected reactors

    NASA Technical Reports Server (NTRS)

    Miley, G. H.

    1978-01-01

    The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters.

  9. A model for calculation of RCS pressure during reflux boiling under reduced inventory conditions and its assessment against PKL data. [Reactor Cooling Systems (RCS)

    SciTech Connect

    Palmrose, D.E. ); Mandl, R.M. )

    1991-01-01

    There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase in the reactor coolant system, and void fraction distributions on the primary side of the system. Mathematical models of these and other physical processes Experiment B4.5.

  10. A citation-based assessment of the performance of U.S. boiling water reactors following extended power up-rates

    NASA Astrophysics Data System (ADS)

    Heidrich, Brenden J.

    Nuclear power plants produce 20 percent of the electricity generated in the U.S. Nuclear generated electricity is increasingly valuable to a utility because it can be produced at a low marginal cost and it does not release any carbon dioxide. It can also be a hedge against uncertain fossil fuel prices. The construction of new nuclear power plants in the U.S. is cautiously moving forward, restrained by high capital costs. Since 1998, nuclear utilities have been increasing the power output of their reactors by implementing extended power up-rates. Power increases of up to 20 percent are allowed under this process. The equivalent of nine large power plants has been added via extended power up-rates. These up-rates require the replacement of large capital equipment and are often performed in concert with other plant life extension activities such as license renewals. This dissertation examines the effect of these extended power up-rates on the safety performance of U.S. boiling water reactors. Licensing event reports are submitted by the utilities to the Nuclear Regulatory Commission, the federal nuclear regulator, for a wide range of abnormal events. Two methods are used to examine the effect of extended power up-rates on the frequency of abnormal events at the reactors. The Crow/AMSAA model, a univariate technique is used to determine if the implementation of an extended power up-rate affects the rate of abnormal events. The method has a long history in the aerospace industry and in the military. At a 95-percent confidence level, the rate of events requiring the submission of a licensing event report decreases following the implementation of an extended power up-rate. It is hypothesized that the improvement in performance is tied to the equipment replacement and refurbishment that is performed as part of the up-rate process. The reactor performance is also analyzed using the proportional hazards model. This technique allows for the estimation of the effects of

  11. Advanced Test Reactor Testing Experience: Past, Present and Future

    SciTech Connect

    Frances M. Marshall

    2005-04-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 48" long and 5.0" diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors -- US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, wherein the target material is placed in a capsule, or plate form, and the capsule is in direct contact with the primary coolant. The next level of complexity of an experiment is an instrumented lead experiment, which allows for active monitoring and control of experiment conditions during the irradiation. The highest level of complexity of experiment is the pressurized water loop experiment, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

  12. Thermal hydraulic analysis for the Oregon State TRIGA reactor using RELAP5-3D

    SciTech Connect

    Marcum, W.R.; Woods, B.G.; Hartman, M.

    2008-07-15

    Thermal hydraulic analyses have being conducted at Oregon State University (OSU) in support of the conversion of the OSU TRIGA reactor (OSTR) core from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel as part of the Reduced Enrichment for Research and Test Reactors program. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the HEU and LEU cores; calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as departure from nuclear boiling ratio (DNBR) for steady state and pulse operation; and perform accident analyses for the accident scenarios identified in the OSTR safety analysis report. RELAP5-3D Version 2.4.2 was implemented to develop a model for the thermal hydraulic study. The OSTR core conversion is planned to take place in late 2008. (author)

  13. Systems analysis of the CANDU 3 Reactor

    SciTech Connect

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H.

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  14. An Overview of the International Reactor Physics Experiment Evaluation Project

    SciTech Connect

    Briggs, J. Blair; Gulliford, Jim

    2014-10-09

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  15. Improvements in Predicting Void Fraction in Subcooled Boiling

    SciTech Connect

    Ha, Kwi Seok; Lee, Yong Bum; No, Hee Cheon

    2005-06-15

    A simple two-phase thermal-hydraulic tool with the drift-flux model has been used to develop a subcooled boiling model. The tool is composed of four governing equations: mixture mass, vapor mass, mixture momentum, and mixture enthalpy. Using the developed tool, various subcooled boiling models were investigated through the published experimental data. In the process of evaluation, two models were developed associated with the subcooled boiling. First, the Saha and Zuber correlation predicting the point of the net vapor generation was modified to consider the thermal and dynamic effects at the high-velocity region. Second, the pumping factor model was developed using the pi-theorem based on parameters related to the bubble generation mechanism, and it produced an additional parameter: the boiling number. The proposed models and several other models were evaluated against a series of subcooled flow boiling experiments at the pressure range of 1 to 146.8 bars. From the root-mean-square analysis for the predicted void fraction in the subcooled boiling region, the results of the proposed model presented the best predictions for the whole-pressure ranges. Also, the implementation of the developed models into RELAP5/MOD3.3 brought about improved results compared to those of the default model of the code.

  16. Two-dimensional computational modeling of sodium boiling in simulated LMFBR fuel-pin bundles

    SciTech Connect

    Dearing, J.F.

    1981-01-01

    Extensive sodium boiling tests have been carried out in two simulated LMFBR fuel pin bundles in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility at Oak Ridge National Laboratory. Experimental results from a 19-pin bundle (THORS Bundle 6A) have been previously reported, and experimental results from a 61-pin bundle (THORS Bundle 9) will be reported soon. The results discussed here are from the 19-pin bundle. Preliminary analysis has shown that the computational methods used and conclusions reached are equally valid for the 61-pin bundle, as well as the 19-pin in-reactor Sodium Loop Safety Facility (SLSF) W-1 experiment. The main result of THORS sodium boiling experimentation is that boiling behavior is determined by two-dimensional effects, i.e., the rates of mass, momentum and energy transfer in the direction perpendicular to the axes of the fuel pins.

  17. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    SciTech Connect

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

  18. Calculated analysis of experiments in fast neutron reactors

    SciTech Connect

    Davydov, V. K. Kalugina, K. M.; Gomin, E. A.

    2012-12-15

    In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.

  19. Calculated analysis of experiments in fast neutron reactors

    NASA Astrophysics Data System (ADS)

    Davydov, V. K.; Kalugina, K. M.; Gomin, E. A.

    2012-12-01

    In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.

  20. The International Reactor Physics Experiment Evaluation Project (IRPhEP)

    SciTech Connect

    Blair Briggs, J.; Sartori, E.; Scott, L.

    2006-07-01

    Since the beginning of the Nuclear Power industry, numerous experiments concerned with nuclear energy and technology have been performed at different research laboratories, worldwide. These experiments required a large investment in terms of infrastructure, expertise, and cost; however, many were performed without a high degree of attention to archival of results for future use. The degree and quality of documentation varies greatly. There is an urgent need to preserve integral reactor physics experimental data, including measurement methods, techniques, and separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. If the data are compromised, it is unlikely that any of these experiments will be repeated again in the future. The International Reactor Physics Evaluation Project (IRPhEP) was initiated, as a pilot activity in 1999 by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. A short history of the IRPhEP is presented and its purposes are discussed in this paper. Accomplishments of the IRPhEP, including the first publication of the IRPhEP Handbook, are highlighted and the future of the project outlined. (authors)

  1. The International Reactor Physics Experiment Evaluation Project (IRPHEP)

    SciTech Connect

    J. Blair Briggs; Enrico Sartori; Lori Scott

    2006-09-01

    Since the beginning of the Nuclear Power industry, numerous experiments concerned with nuclear energy and technology have been performed at different research laboratories, worldwide. These experiments required a large investment in terms of infrastructure, expertise, and cost; however, many were performed without a high degree of attention to archival of results for future use. The degree and quality of documentation varies greatly. There is an urgent need to preserve integral reactor physics experimental data, including measurement methods, techniques, and separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. If the data are compromised, it is unlikely that any of these experiments will be repeated again in the future. The International Reactor Physics Evaluation Project (IRPhEP) was initiated, as a pilot activity in 1999 by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. A short history of the IRPhEP is presented and its purposes are discussed in this paper. Accomplishments of the IRPhEP, including the first publication of the IRPhEP Handbook, are highlighted and the future of the project outlined.

  2. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    NASA Astrophysics Data System (ADS)

    Abe, K.; Kohyama, A.; Namba, C.; Wiffen, F. W.; Jones, R. H.

    1998-10-01

    A Japan-USA Program of irradiation experiments for fusion research, "JUPITER", has been established as a 6 year program from 1995 to 2000. The goal is to study "the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment". This is phase-three of the collaborative program, which follows RTNS-II Program (Phase-1: 1982-1986) and FFTF/MOTA Program (Phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA Program, JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects.

  3. Observation of reactor electron antineutrinos disappearance in the RENO experiment.

    PubMed

    Ahn, J K; Chebotaryov, S; Choi, J H; Choi, S; Choi, W; Choi, Y; Jang, H I; Jang, J S; Jeon, E J; Jeong, I S; Joo, K K; Kim, B R; Kim, B C; Kim, H S; Kim, J Y; Kim, S B; Kim, S H; Kim, S Y; Kim, W; Kim, Y D; Lee, J; Lee, J K; Lim, I T; Ma, K J; Pac, M Y; Park, I G; Park, J S; Park, K S; Shin, J W; Siyeon, K; Yang, B S; Yeo, I S; Yi, S H; Yu, I

    2012-05-11

    The RENO experiment has observed the disappearance of reactor electron antineutrinos, consistent with neutrino oscillations, with a significance of 4.9 standard deviations. Antineutrinos from six 2.8  GW(th) reactors at the Yonggwang Nuclear Power Plant in Korea, are detected by two identical detectors located at 294 and 1383 m, respectively, from the reactor array center. In the 229 d data-taking period between 11 August 2011 and 26 March 2012, the far (near) detector observed 17102 (154088) electron antineutrino candidate events with a background fraction of 5.5% (2.7%). The ratio of observed to expected numbers of antineutrinos in the far detector is 0.920±0.009(stat)±0.014(syst). From this deficit, we determine sin(2)2θ(13)=0.113±0.013(stat)±0.019(syst) based on a rate-only analysis. PMID:23003027

  4. Benchmark experiments for validation of reaction rates determination in reactor dosimetry

    NASA Astrophysics Data System (ADS)

    Rataj, J.; Huml, O.; Heraltova, L.; Bily, T.

    2014-11-01

    The precision of Monte Carlo calculations of quantities of neutron dosimetry strongly depends on precision of reaction rates prediction. Research reactor represents a very useful tool for validation of the ability of a code to calculate such quantities as it can provide environments with various types of neutron energy spectra. Especially, a zero power research reactor with well-defined core geometry and neutronic properties enables precise comparison between experimental and calculated data. Thus, at the VR-1 zero power research reactor, a set of benchmark experiments were proposed and carried out to verify the MCNP Monte Carlo code ability to predict correctly the reaction rates. For that purpose two frequently used reactions were chosen: He-3(n,p)H-3 and Au-197(n,γ)Au-198. The benchmark consists of response measurement of small He-3 gas filled detector in various positions of reactor core and of activated gold wires placed inside the core or to its vicinity. The reaction rates were calculated in MCNP5 code utilizing a detailed model of VR-1 reactor which was validated for neutronic calculations at the reactor. The paper describes in detail the experimental set-up of the benchmark, the MCNP model of the VR-1 reactor and provides a comparison between experimental and calculated data.

  5. Numerical experiments on evaporation and explosive boiling of ultra-thin liquid argon film on aluminum nanostructure substrate

    NASA Astrophysics Data System (ADS)

    Wang, Weidong; Zhang, Haiyan; Tian, Conghui; Meng, Xiaojie

    2015-04-01

    Evaporation and explosive boiling of ultra-thin liquid film are of great significant fundamental importance for both science and engineering applications. The evaporation and explosive boiling of ultra-thin liquid film absorbed on an aluminum nanostructure solid wall are investigated by means of molecular dynamics simulations. The simulated system consists of three regions: liquid argon, vapor argon, and an aluminum substrate decorated with nanostructures of different heights. Those simulations begin with an initial configuration for the complex liquid-vapor-solid system, followed by an equilibrating system at 90 K, and conclude with two different jump temperatures, including 150 and 310 K which are far beyond the critical temperature. The space and time dependences of temperature, pressure, density number, and net evaporation rate are monitored to investigate the phase transition process on a flat surface with and without nanostructures. The simulation results reveal that the nanostructures are of great help to raise the heat transfer efficiency and that evaporation rate increases with the nanostructures' height in a certain range.

  6. Numerical experiments on evaporation and explosive boiling of ultra-thin liquid argon film on aluminum nanostructure substrate.

    PubMed

    Wang, Weidong; Zhang, Haiyan; Tian, Conghui; Meng, Xiaojie

    2015-01-01

    Evaporation and explosive boiling of ultra-thin liquid film are of great significant fundamental importance for both science and engineering applications. The evaporation and explosive boiling of ultra-thin liquid film absorbed on an aluminum nanostructure solid wall are investigated by means of molecular dynamics simulations. The simulated system consists of three regions: liquid argon, vapor argon, and an aluminum substrate decorated with nanostructures of different heights. Those simulations begin with an initial configuration for the complex liquid-vapor-solid system, followed by an equilibrating system at 90 K, and conclude with two different jump temperatures, including 150 and 310 K which are far beyond the critical temperature. The space and time dependences of temperature, pressure, density number, and net evaporation rate are monitored to investigate the phase transition process on a flat surface with and without nanostructures. The simulation results reveal that the nanostructures are of great help to raise the heat transfer efficiency and that evaporation rate increases with the nanostructures' height in a certain range. PMID:25918494

  7. Shifts of neutrino oscillation parameters in reactor antineutrino experiments with non-standard interactions

    NASA Astrophysics Data System (ADS)

    Li, Yu-Feng; Zhou, Ye-Ling

    2014-11-01

    We discuss reactor antineutrino oscillations with non-standard interactions (NSIs) at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline) dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO) as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters. Long baseline reactor antineutrino experiments, such as KamLAND [10,11]. The aim of these experiments is to observe the slow oscillation with Δ21 and measure the corresponding oscillation parameters Δm212 and θ12. Short baseline reactor antineutrino experiments, such as Daya Bay [1-3], Double CHOOZ [4], RENO [5]. They are designed to observe the fast oscillation with Δ31 and Δ32 (or equivalently, Δee[3]) and measure the corresponding oscillation parameters Δmee2, θ13. Medium baseline reactor antineutrino experiments. They stand for the next generation experiments of reactor antineutrinos, with typical representatives of Jiangmen Underground Neutrino Observatory (JUNO) [12] and RENO-50 [13]. They can determine the neutrino mass ordering (m13 or m3reactor antineutrino experiments. High-dimensional operators originating from new physics can contribute to the neutrino oscillation in the form of non-standard interactions (NSIs) [14

  8. Slow control systems of the Reactor Experiment for Neutrino Oscillation

    NASA Astrophysics Data System (ADS)

    Choi, J. H.; Jang, H. I.; Choi, W. Q.; Choi, Y.; Jang, J. S.; Jeon, E. J.; Joo, K. K.; Kim, B. R.; Kim, H. S.; Kim, J. Y.; Kim, S. B.; Kim, S. Y.; Kim, W.; Kim, Y. D.; Ko, Y. J.; Lee, J. K.; Lim, I. T.; Pac, M. Y.; Park, I. G.; Park, J. S.; Park, R. G.; Seo, H. K.; Seo, S. H.; Shin, C. D.; Siyeon, K.; Yeo, I. S.; Yu, I.

    2016-02-01

    The RENO experiment has been in operation since August 2011 to measure reactor antineutrino disappearance using identical near and far detectors. For accurate measurements of neutrino mixing parameters and efficient data taking, it is crucial to monitor and control the detector in real time. Environmental conditions also need to be monitored for stable operation of detectors as well as for safety reasons. In this paper, we report the design, hardware, operation, and performance of the slow control system.

  9. Numerical Simulation of Boiling Two-Phase Flow in Tight-Lattice Rod Bundle by 3-Dimensional Two-Fluid Model Code ACE-3D

    NASA Astrophysics Data System (ADS)

    Yoshida, Hiroyuki; Misawa, Takeharu; Takase, Kazuyuki

    Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase flow simulation method such as interface tracking method or particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D that adopts boundary fitted coordinate system in order to simulate complex shape flow channel. In this paper, boiling two-phase flow analysis in a tight-lattice rod bundle was performed by ACE-3D code. The parallel computation using 126 CPUs was applied to this analysis. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. The tendency of void fraction distribution agreed with the measurement results by neutron radiography qualitatively. To evaluate effects of two-phase flow model used in ACE-3D code, numerical simulation of boiling two-phase in tight-lattice rod bundle with no lift force model was also performed. From the comparison of calculated results, it was concluded that the effects of lift force model were not so large for overall void fraction distribution of tight-lattice rod bundle. However, the lift force model is important for local void fraction distribution of fuel bundles.

  10. Development of heat transfer enhancement techniques for external cooling of an advanced reactor vessel

    NASA Astrophysics Data System (ADS)

    Yang, Jun

    Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (<800 °C). With the upgrade and development of advanced power reactors, however, enhancing the nucleate boiling rate and its upper limit, Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub

  11. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    SciTech Connect

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  12. Bounds on sterile neutrino parameters from reactor experiments

    NASA Astrophysics Data System (ADS)

    Bora, Kalpana

    2014-03-01

    In this work, we present a realistic analysis of the potential of the present-day reactor experiments Double Chooz, Daya Bay and RENO for probing the existence of sterile neutrinos. We present exclusion regions for sterile oscillation parameters and find that these experimental set-ups give significant bounds on the parameter Θee especially in the low sterile oscillation region 0.01 < Δm241 < 0.05 eV2. These bounds can add to our understanding of the sterile neutrino sector since there is still a tension in the allowed regions from different experiments for sterile parameters.

  13. Nucleate boiling pressure drop in an annulus: Book 5

    SciTech Connect

    Not Available

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux. This document consists solely of the plato file index from 11/87 to 11/90.

  14. Development of the Packed Bed Reactor ISS Flight Experiment

    NASA Technical Reports Server (NTRS)

    Patton, Martin O.; Bruzas, Anthony E.; Rame, Enrique; Motil, Brian J.

    2012-01-01

    Packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a leading candidate as a potential unit operation in support of long duration human space exploration. On earth, this type of reactor accounts for approximately 80% of all the reactors used in the chemical process industry today. Development of this technology for space exploration is truly crosscutting with many other potential applications (e.g., in-situ chemical processing of planetary materials and transport of nutrients through soil). NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. The first model developed by Motil et al., (2003) is based on a modified Ergun equation. This model was demonstrated at moderate gas and liquid flow rates, but extension to the lower flow rates expected in many advanced life support systems must be validated. The other model, developed by Guo et al., (2004) is based on Darcy s (1856) law for two-phase flow. This model has been validated for a narrow range of flow parameters indirectly (without full instrumentation) and included test points where the flow was not fully developed. The flight experiment presented will be designed with removable test sections to test the hydrodynamic models. The experiment will provide flexibility to test additional beds with different types of packing in the future. One initial test bed is based on the VRA (Volatile Removal Assembly), a packed bed reactor currently on ISS whose behavior in micro-gravity is not fully understood. Improving the performance of this system through an accurate model will increase our ability to purify water in the space environment.

  15. R1234yf vs. R134a Flow Boiling Heat Transfer Inside a 3.4 mm ID Microfin Tube

    NASA Astrophysics Data System (ADS)

    Diani, A.; Mancin, S.; Rossetto, L.

    2014-11-01

    The refrigerant charge minimization as well as the use of eco-friendly fluids can be considered two of the most important targets for these applications to cope with the new environmental challenges. This paper compares the R1234yf and R134a flow boiling heat transfer and pressure drop measurements inside a small microfin tube with internal diameter at the fin tip of 3.4 mm. This study is carried out in an experimental facility built at the Dipartimento di Ingegneria Industriale of the University of Padova especially designed to study both single and two phase heat transfer processes. The microfin tube is brazed inside a copper plate and electrically heated from the bottom. Several T -type thermocouples are inserted in the wall to measure the temperature distribution during the phase change process. In particular, the experimental measurements were carried out at constant saturation temperature of 30 °C, by varying the refrigerant mass velocity between 190 kg m-2 s-1 and 940 kg m-2 s-1, the vapour quality from 0.2 to 0.99, at different imposed heat fluxes. The two refrigerants are compared considering the values of the two-phase heat transfer coefficient and pressure drop.

  16. The detector system of the Daya Bay reactor neutrino experiment

    DOE PAGESBeta

    An, F. P.

    2015-12-15

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 22θ13 and the effective mass splitting Δm2ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrinomore » mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.« less

  17. The detector system of the Daya Bay reactor neutrino experiment

    SciTech Connect

    An, F. P.

    2015-12-15

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 213 and the effective mass splitting Δm2ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  18. The detector system of the Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    An, F. P.; Bai, J. Z.; Balantekin, A. B.; Band, H. R.; Beavis, D.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. L.; Butorov, I.; Cao, D.; Cao, G. F.; Cao, J.; Carr, R.; Cen, W. R.; Chan, W. T.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chasman, C.; Chen, H. Y.; Chen, H. S.; Chen, M. J.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, X. S.; Chen, Y. X.; Chen, Y.; Cheng, J. H.; Cheng, J.; Cheng, Y. P.; Cherwinka, J. J.; Chidzik, S.; Chow, K.; Chu, M. C.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dong, L.; Dove, J.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fang, S. D.; Fu, J. Y.; Fu, Z. W.; Ge, L. Q.; Ghazikhanian, V.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gornushkin, Y. A.; Grassi, M.; Greenler, L. S.; Gu, W. Q.; Guan, M. Y.; Guo, R. P.; Guo, X. H.; Hackenburg, R. W.; Hahn, R. L.; Han, R.; Hans, S.; He, M.; He, Q.; He, W. S.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hinrichs, P.; Ho, T. H.; Hoff, M.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. M.; Hu, L. J.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. Z.; Huang, H. X.; Huang, P. W.; Huang, X.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiang, H. J.; Jiang, W. Q.; Jiao, J. B.; Johnson, R. A.; Joseph, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, C. Y.; Lai, W. C.; Lai, W. H.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, M. K. P.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Lewis, C. A.; Li, B.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, J.; Li, N. Y.; Li, Q. J.; Li, S. F.; Li, S. C.; Li, W. D.; Li, X. B.; Li, X. N.; Li, X. Q.; Li, Y.; Li, Y. F.; Li, Z. B.; Liang, H.; Liang, J.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. X.; Lin, S. K.; Lin, Y. C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, B. J.; Liu, C.; Liu, D. W.; Liu, H.; Liu, J. L.; Liu, J. C.; Liu, S.; Liu, S. S.; Liu, X.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, A.; Luk, K. B.; Luo, T.; Luo, X. L.; Ma, L. H.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Mayes, B.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Monari Kebwaro, J.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Nie, Y. B.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pagac, A.; Pan, H.-R.; Patton, S.; Pearson, C.; 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.; Sands, W. R.; Seilhan, B.; Shao, B. B.; Shih, K.; Song, W. Y.; Steiner, H.; Stoler, P.; Stuart, M.; Sun, G. X.; Sun, J. L.; Tagg, N.; Tam, Y. H.; Tanaka, H. K.; Tang, W.; Tang, X.; Taychenachev, D.; Themann, H.; Torun, Y.; Trentalange, S.; Tsai, O.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viaux, N.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, T.; Wang, W.; Wang, W. W.; Wang, X. T.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Wenman, D. L.; Whisnant, K.; White, C. G.; Whitehead, L.; Whitten, C. A.; Wilhelmi, J.; Wise, T.; Wong, H. C.; Wong, H. L. H.; Wong, J.; Wong, S. C. F.; Worcester, E.; Wu, F. F.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xiang, S. T.; Xiao, Q.; Xing, Z. Z.; Xu, G.; Xu, J. Y.; Xu, J. L.; Xu, J.; Xu, W.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Yip, K.; Young, B. L.; Yu, G. Y.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, H. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. X.; Zhang, Q. M.; Zhang, S. H.; Zhang, X. T.; Zhang, Y. C.; Zhang, Y. H.; 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. F.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhou, Z. Y.; Zhuang, H. L.; Zimmerman, S.; Zou, J. H.

    2016-03-01

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of νbare oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin2 2θ13 and the effective mass splitting Δ mee2. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors' baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This paper describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  19. Status of the Daya Bay Reactor Neutrino Oscillation Experiment

    SciTech Connect

    Daya Bay Collaboration; Lin, Cheng-Ju Stephen

    2010-12-15

    The last unknown neutrino mixing angle theta_13 is one of the fundamental parameters of nature; it is also a crucial parameter for determining the sensitivity of future long-baseline experiments aimed to study CP violation in the neutrino sector. Daya Bay is a reactor neutrino oscillation experiment designed to achieve a sensitivity on the value of sin^2(2*theta_13) to better than 0.01 at 90percent CL. The experiment consists of multiple identical detectors placed underground at different baselines to minimize systematic errors and suppress cosmogenic backgrounds. With the baseline design, the expected anti-neutrino signal at the far site is about 360 events per day and at each of the near sites is about 1500 events per day. An overview and current status of the experiment will be presented.

  20. Boiling Point

    NASA Technical Reports Server (NTRS)

    Jansen, Michael C.

    2002-01-01

    The author recounts his experiences he helped to investigate the accident which destroyed the Space Shuttle Challenger. The focus was on how he used novel approaches to investigate heat transfer in the shuttle's hydrogen tank, after an expert he sought for advice proved unhelpful.

  1. Beryllium reflected cavity reactor for UF6 critical experiments

    NASA Technical Reports Server (NTRS)

    Jarvis, G. A.; Bernard, W.; Helmick, H. H.; White, R.

    1975-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diam by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials are available. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  2. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    SciTech Connect

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  3. Predicting Activation of Experiments Inside the Annular Core Research Reactor

    SciTech Connect

    Greenberg, Joseph Isaac

    2015-11-01

    The objective of this thesis is to create a program to quickly estimate the radioactivity and decay of experiments conducted inside of the Annular Core Research Reactor at Sandia National Laboratories and eliminate the need for users to write code. This is achieved by model the neutron fluxes in the reactor’s central cavity where experiments are conducted for 4 different neutron spectra using MCNP. The desired neutron spectrum, experiment material composition, and reactor power level are then input into CINDER2008 burnup code to obtain activation and decay information for every isotope generated. DREAD creates all of the files required for CINDER2008 through user selected inputs in a graphical user interface and executes the program for the user and displays the resulting estimation for dose rate at various distances. The DREAD program was validated by weighing and measuring various experiments in the different spectra and then collecting dose rate information after they were irradiated and comparing it to the dose rates that DREAD predicted. The program provides results with an average of 17% higher estimates than the actual values and takes seconds to execute.

  4. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    NASA Astrophysics Data System (ADS)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  5. 75 FR 32229 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on ABWR

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-07

    ... Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on June 23-24, 2010, Room T-2B3... were published in the Federal Register on October 14, 2009 (74 FR 58268-58269). Detailed...

  6. Comparative studies of pool boiling heat transfer with nano-fluids on porous surface

    NASA Astrophysics Data System (ADS)

    Niu, Gengwen; Li, Ji

    2015-12-01

    Characteristics and regime of pool boiling on copper porous coating and polished surface are studied experimentally for two working medium (Al2O3 and CuO nano-fluids). The highlights of this work are that the combination of nano-fluids and porous surface is used to explore the heat transfer performance and the conducted visual experiments can provide an insight in the enhancement of boiling heat transfer with this combination.

  7. The Muon System of the Daya Bay Reactor Antineutrino Experiment

    DOE PAGESBeta

    An, F. P.; Hackenburg, R. W.; Brown, R. E.; Chasman, C.; Dale, E.; Diwan, M. V.; Gill, R.; Hans, S.; Isvan, Z.; Jaffe, D. E.; et al

    2014-10-05

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described. (auth)

  8. MHTGR: New production reactor summary of experience base

    SciTech Connect

    Not Available

    1988-03-01

    Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs.

  9. The muon system of the Daya Bay Reactor antineutrino experiment

    NASA Astrophysics Data System (ADS)

    An, F. P.; Balantekin, A. B.; Band, H. R.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. E.; Butorov, I.; Cao, G. F.; Cao, J.; Carr, R.; Chan, Y. L.; Chang, J. F.; Chang, L.; Chang, Y.; Chasman, C.; Chen, H. S.; Chen, H. Y.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, Y.; Chen, Y. X.; Cheng, Y. P.; Cherwinka, J. J.; Chu, M. C.; Cummings, J. P.; Dale, E.; de Arcos, J.; Deng, Z. Y.; Ding, Y. Y.; Diwan, M. V.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fu, J. Y.; Ge, L. Q.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gu, W. Q.; Guan, M. Y.; Guo, X. H.; Hackenburg, R. W.; Han, G. H.; Hans, S.; He, M.; He, Q.; Heeger, K. M.; Heng, Y. K.; Hinrichs, P.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. J.; Hu, L. M.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, H. Z.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jetter, S.; Ji, X. L.; Ji, X. P.; Jiang, H. J.; Jiao, J. B.; Johnson, R. A.; Kang, L.; Kebwaro, J. M.; Kettell, S. H.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, W. C.; Lai, W. H.; Lau, K.; Lebanowski, L.; Lee, J.; Lei, R. T.; Leitner, R.; Leung, A.; Leung, J. K. C.; Lewis, C. A.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. K.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, H.; Liu, J. C.; Liu, J. L.; Liu, S. S.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Luk, K. B.; Ma, Q. M.; Ma, X. B.; Ma, X. Y.; Ma, Y. Q.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Nemchenok, I.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevski, A.; Patton, S.; Pec, V.; Pearson, C. E.; 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.; Tam, Y. H.; Tang, X.; Themann, H.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, W. W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wilhelmi, J.; 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, G. H.; Xu, J.; Xu, J. L.; Xu, J. Y.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Young, B. L.; Yu, G. Y.; Yu, J. Y.; Yu, Z. Y.; Zang, S. L.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. M.; Zhang, S. H.; Zhang, Y. H.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Z. J.; Zhang, Z. P.; Zhang, Z. Y.; Zhao, J.; Zhao, Q. W.; Zhao, Y.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, Z. Y.; Zhuang, H. L.; Zou, J. H.

    2015-02-01

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described.

  10. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    SciTech Connect

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  11. How Historical Experiments Can Improve Scientific Knowledge and Science Education: The Cases of Boiling Water and Electrochemistry

    ERIC Educational Resources Information Center

    Chang, Hasok

    2011-01-01

    I advance some novel arguments for the use of historical experiments in science education. After distinguishing three different types of historical experiments and their general purposes, I define "complementary experiments", which can recover lost scientific knowledge and extend what has been recovered. Complementary experiments can help science…

  12. Startup experience at the University of Texas TRIGA Mark II Reactor

    SciTech Connect

    Bauer, Thomas L.; Wehring, Bernard W.

    1992-07-01

    After eight years of singular effort, the UT-TRIGA Mark II research reactor was licensed and is fully operational. This reactor is the focus of a new reactor laboratory facility which is located at the Balcones Research Center, a north Austin campus of The University of Texas at Austin. The UT-TRIGA reactor is licensed for 1.1 MW steady power operation and 3 dollar pulsing. A startup program was implemented upon receipt of the facility license on January 17, 1992. Several facility features are unique to this startup. Among these were the use of fuel with various burnup and a digital control system. The reactor laboratory staff with assistance from a General Atomics instrumentation engineer performed all phases of the startup program. Core loading began in February 1992 with final testing completed in May 1992. Several unusual problems were encountered during this time. Experiment authorizations have been written to resume Neutron Activation Analysis programs and isotope production. Several neutron beam tube experiments are in the design and test phase. (author)

  13. Flow excursion experiments with a production reactor assembly mockup

    SciTech Connect

    Rush, G.C.; Blake, J.E. ); Nash, C.A. )

    1990-01-01

    A series of power ramp and loss-of-coolant accidents were simulated with an electrically heated mockup of a Savannah River Site production reactor assembly. The one-to-one scale mockup had full multichannel annular geometry in its heated section in addition to prototypical inlet and outlet endfitting hardware. Power levels causing void generation and flow instability in the water coolant flowing through the mockup were found under different transient and quasi-steady state test conditions. A reasonably sharp boundary between initial operating powers leading to or not leading to flow instability were found: that being 0.2 MW or less on power levels of 4 to 6.3 MW. Void generation occurred before, but close to, the point of flow instability. The data were taken in support of the Savannah River reactor limits program and will be used in continuing code benchmarking efforts. 6 refs., 12 figs., 2 tabs.

  14. Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

    PubMed

    van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

    2007-10-01

    The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed. PMID:17583763

  15. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    SciTech Connect

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  16. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    SciTech Connect

    Brumovsky, M.; Polachova, H.

    1995-11-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber`s, Hardrath-Ohman`s as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared.

  17. THORAX pretest prediction of a sodium-boiling transient in a 19-pin simulated LMFBR driver bundle

    SciTech Connect

    Rose, S.D.

    1982-01-01

    Experiments will be conducted in the Thermal-Hydraulic Out-of-Reactor Safety-Shutdown Heat Removal System (THORS-SHRS) Assembly 1 loop at Oak Ridge National Laboratory (ORNL) to model the behavior of a reactor during degraded decay heat removal conditions. The test section is to consist of two parallel 19-pin electrically-heated driver bundles, typical of U.S. Large Developmental Plant (LDP) Liquid Metal Fast Breeder Reactor (LMFBR) design. Analysis of these experiments will include using THORAX, a two-dimensional boiling model which assumes an equilibrium mixture two-phase flow (with slip). A THORAX prediction is presented for a single-bundle forced convection boiling-to-dryout transient at 15.8 kW/pin.

  18. Rod-bundle transient-film boiling of high-pressure water in the liquid-deficient regime. [PWR

    SciTech Connect

    Morris, D.G.; Mullins, C.B.; Yoder, G.L.

    1982-01-01

    Results are reported from a recent experiment investigating dispersed flow film boiling of high pressure water in upflow through a rod bundle. The data, obtained under mildly transient conditions, are used to assess correlations currently used to predict heat transfer in these circumstances. In light of the scarcity of similar data, the data should prove useful in the development and assessment of new heat transfer models. The experiment was conducted at the Oak Ridge National Laboratory in the Thermal-Hydraulic Test Facility, a highly instrumented, non-nuclear, pressurized-water loop containing 64, 3.66-m (12-ft) long rods (of which 60 are electrically heated). The rods are arranged in a square array typical of 17 x 17 fuel rod assemblies in late generation PWRs. Data were collected over typical reactor blowdown parameter ranges.

  19. Thermal-hydraulic aspects of flow inversion in a research reactor

    SciTech Connect

    Smith, R.S.; Woodruff, W.L.

    1986-11-01

    PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.

  20. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    SciTech Connect

    Vierow, Karen

    2008-09-26

    This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.

  1. Resolving the reactor neutrino anomaly with the KATRIN neutrino experiment

    NASA Astrophysics Data System (ADS)

    Formaggio, J. A.; Barrett, J.

    2011-11-01

    The KArlsruhe TRItium Neutrino experiment (KATRIN) combines an ultra-luminous molecular tritium source with an integrating high-resolution spectrometer to gain sensitivity to the absolute mass scale of neutrinos. The projected sensitivity of the experiment on the electron neutrino mass is 200 meV at 90% C.L. With such unprecedented resolution, the experiment is also sensitive to physics beyond the Standard Model, particularly to the existence of additional sterile neutrinos at the eV mass scale. A recent analysis of available reactor data appears to favor the existence of such a sterile neutrino with a mass splitting of | Δmsterile | 2 ⩾ 1.5eV2 and mixing strength of sin2 2θsterile = 0.17 ± 0.08 at 95% C.L. Upcoming tritium beta decay experiments should be able to rule out or confirm the presence of the new phenomenon for a substantial fraction of the allowed parameter space.

  2. Investigations of Mechanisms Associated with Nucleate Boiling Under Microgravity Conditions

    NASA Technical Reports Server (NTRS)

    Dhir, Vijay K.; Hasan, M.; Chao, David (Technical Monitor)

    2003-01-01

    In this work a building block type of approach is used so that a basic understanding of the processes that contribute to nucleate boiling heat fluxes under microgravity conditions can be developed. This understanding will lead to development of a mechanistic model for nucleate boiling heat transfer which could eventually be used as a design tool in space applications. Task Description Task 1: Fabrication of the Experimental Setup. Under this task, the test section and liquid holding and viewing chambers will be fabricated. Artificial cylinder cavities will be formed on silicon wafers. A single cavity and two or four cavities, with a prescribed spacing and size, will be formed. The desired nucleation wall superheat will be used to determine the size of the mouth of the cavities. Task 2: Experiments. The basic experiments for flow and temperature field around single and multiple (2 or 4 separated or merged bubbles growing on downward facing or inclined surfaces) will be carried out under normal gravity conditions. The experiments will be conducted at one atmosphere pressure, but liquid subcooling will be varied from 0 to 30C. Water and PF-5050 will be used as test liquids. Task 3: Analytical/Numerical Models. In this task, transient temperature and flow field in vapor and liquid will be determined during growth of a single bubble. Analysis will include the evolution of the vapor-liquid interface and development of microlayer underneath the bubbles. For merged bubbles, detailed calculations of flow and temperature field will be carried out for transient shapes of vapor stems supporting a large bubble and the corresponding evaporation rate. Flow and temperature field for a bubble sliding along a heated wall will also be determined. Microgravity conditions will be simulated and a framework of a numerical tool for prediction of nucleate boiling heat fluxes under microgravity conditions will be developed. Task 4: Experiments in a KC-135. To understand bubble growth and

  3. Helium-3 blankets for tritium breeding in fusion reactors

    NASA Technical Reports Server (NTRS)

    Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

    1988-01-01

    It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

  4. Nucleate pool boiling in the long duration low gravity environment of the space shuttle

    NASA Technical Reports Server (NTRS)

    Hasan, M. M.; Lin, C. S.; Knoll, R. H.; Bentz, M. D.; Meserole, J. S.

    1993-01-01

    The results are presented of an experimental study of nucleate pool boiling performed in the low gravity environment of the space shuttle. Photographic observations of pool boiling in Freon 113 were obtained during the 'Tank Pressure Control Experiment', flown on the Space Transportation System STS-43 in August 1991. Nucleate boiling data from large (relative to bubble size) flat heating surfaces (0.1046 by 0.0742 m) was obtained at very low heat fluxes (0.22 to 1.19 kw/so m). The system pressure and the bulk liquid subcooling varied in the range of 40 to 60 kPa and 3 to 5 C respectively. Thirty-eight boiling tests, each of 10 min duration for a given heat flux, were conducted. Measurements included the heater power, heater surface temperature, the liquid temperature and the system pressure as functions of heating time. Video data of the first 2 min of heating was recorded for each test. In some tests the video clearly shows the inception of boiling and the growth and departure of bubbles from the surface during the first 2 min of heating. In the absence of video data, the heater temperature variation during heating shows the inception of boiling and stable nucleate boiling. During the stable nucleate boiling, the wall superheat varied between 2.8 to 3.8 C for heat fluxes in the range of 0.95 to 1.19 kw/so m. The wall superheat at the inception of boiling varied between 2 to 13 C.

  5. Nucleate pool boiling in the long duration low gravity environment of the Space Shuttle

    NASA Technical Reports Server (NTRS)

    Hasan, M. M.; Lin, C. S.; Knoll, R. H.; Bentz, M. D.; Meserole, J. S.

    1993-01-01

    The results are presented of an experimental study of nucleate pool boiling performed in the low gravity environment of the space shuttle. Photographic observations of pool boiling in Freon 113 were obtained during the 'Tank Pressure Control Experiment,' flown on the Space Transportation System, STS-43 in August 1991. Nucleate boiling data from large (relative to bubble size) flat heating surfaces (0.1046 by 0.0742 m) was obtained at very low heat fluxes (0.22 to 1.19 kW/sq m). The system pressure and the bulk liquid subcooling varied in the range of 40 to 60 kPa and 3 to 5 C respectively. Thirty-eight boiling tests, each of 10-min duration for a given heat flux, were conducted. Measurements included the heater power, heater surface temperature, the liquid temperature and the system pressure as functions of heating time. Video data of the first 2 min of heating was recorded for each test. In some tests the video clearly shows the inception of boiling and the growth and departure of bubbles from the surface during the first 2 min of heating. In the absence of video data, the heater temperature variation during heating shows the inception of boiling and stable nucleate boiling. During the stable nucleate boiling, the wall superheat varied between 2.8 to 3.8 C for heat fluxes in the range of 0.95 to 1.19 kW/sq m. The wall superheat at the inception of boiling varied between 2 to 13 C.

  6. Analysis of epithermal neutron beam experiments at the HIFAR reactor

    SciTech Connect

    Harrington, B.V.; Constantine, G.

    1995-01-01

    A calculated model of the entire core of the DIDO class reactor HIFAR has been used to analyze epithermal neutron beam experiments. In the experiments, an off-center fuel element was replaced by a dummy fuel element voided by a dry liner in which an aluminium spectrum shifter was suspended at core center to extract the beam. Various combinations of the filter materials aluminum, iron, sulfur, titanium, and cadmium were inserted near the top of the dry liner, and liquid argon was placed in a cryostat above the dummy element. Reaction rates were measured in a fission chamber, sandwiched between various thicknesses of polyethylene, in order to assess the accuracy of the calculational model for different regions of the neutron energy spectrum of the beam. The neutron source distribution of the HIFAR core was obtained from a three-dimensional diffusion calculation, with burnup-dependent fuel compositions and fission products included, using the AUS modular code scheme. Argon cross sections were generated from ENDL-84 data and resonance parameters taken from Neutron Cross Sections (1984). A whole-core MCNP source calculation was used to analyze the experiments giving good agreement between measured and calculated reaction rates. This whole-core model of HIFAR may be applied with confidence to predict the performance of filtered beams for boron neutron capture therapy and also to other HIFAR calculations.

  7. Evaporation, Boiling and Bubbles

    ERIC Educational Resources Information Center

    Goodwin, Alan

    2012-01-01

    Evaporation and boiling are both terms applied to the change of a liquid to the vapour/gaseous state. This article argues that it is the formation of bubbles of vapour within the liquid that most clearly differentiates boiling from evaporation although only a minority of chemistry textbooks seems to mention bubble formation in this context. The…

  8. Nucleate boiling pressure drop in an annulus: Book 8

    SciTech Connect

    Not Available

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of tables of temperature measurements.

  9. Nucleate boiling pressure drop in an annulus: Book 4

    SciTech Connect

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements.

  10. Nucleate boiling pressure drop in an annulus: Book 2

    SciTech Connect

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux.

  11. Pool boiling from rotating and stationary spheres in liquid nitrogen

    NASA Technical Reports Server (NTRS)

    Cuan, Winston M.; Schwartz, Sidney H.

    1988-01-01

    Results are presented for a preliminary experiment involving saturated pool boiling at 1 atm from rotating 2 and 3 in. diameter spheres which were immersed in liquid nitrogen (LN2). Additional results are presented for a stationary, 2 inch diameter sphere, quenched in LN2, which were obtained utilizing a more versatile and complete experimental apparatus that will eventually be used for additional rotating sphere experiments. The speed for the rotational tests was varied from 0 to 10,000 rpm. The stationary experiments parametrically varied pressure and subcooling levels from 0 to 600 psig and from 0 to 50 F, respectively. During the rotational tests, a high speed photographic analysis was undertaken to measure the thickness of the vapor film surrounding the sphere. The average Nusselt number over the cooling period was plotted against the rotational Reynolds number. Stationary sphere results included local boiling heat transfer coefficients at different latitudinal locations, for various pressure and subcooling levels.

  12. Boiling Performance of Antifreeze Solutions in a Saturate Pool Boiling System

    NASA Astrophysics Data System (ADS)

    Matsumura, Kunihito; Kaminaga, Fumito

    Nucleate boiling of binary mixtures is of particular importance in a various industries. The purpose of the present study is to provide experimental data and prediction method for nucleate boiling heat transfer of anti-freeze solutions, Propylene-glycol (PG)/water and Ethylene-glycol (EG)/water. The pool nucleate boiling experiments were carried out under a saturated and atmospheric condition. The platinum wire of 0.3 mm diameter was used as the heating surface. The mole fractions of solutions are varied from 0.85 to 1. It was found that the heat transfer coefficient gradually decreases with increasing fraction of anti-freeze to water. It was also shown that a small addition of propylene-glycol and ethylene-glycol also decreases the CHF value far below that of pure water. It is concluded that the correlation proposed by Fujita for several binary mixtures can well predict the heat transfer coefficients within almost ±5% accuracy for every concentration of present anti-freeze solutions.

  13. Characterization of Monodispersed γ-Al2O3 Particles, Synthesized by Homogeneous Precipitation under Reflux Boiling

    NASA Astrophysics Data System (ADS)

    Haq, Ikram Ul; Akhtar, Khalida; Khan, Zia Ullah

    2015-07-01

    In this study, micron size uniform particles of aluminum compounds were synthesized by the homogeneous precipitation method under reflux boiling, using aluminum nitrate, urea and potassium sulfate as starting materials. In this method, varying amounts of the reactants were heated in a Pyrex glass round bottom flask at the boiling temperature for various periods of time, which led to the production of precipitated particles in the reactant mixtures. Scanning electron microscopic (SEM) examination of the precipitated particles revealed that uniformity in particle morphology of the precipitated particles was dependent on the applied experimental conditions. As such, the latter were extensively optimized for obtaining particles of uniform morphology and size. Some of the batches of the precipitated particles were calcined at the elevated temperature (800 and 1000 °C) for transformation into crystalline solids. Similarly, the as prepared and calcined particles were characterized by SEM, XRD, TGA/DTA and FTIR. XRD results revealed that at 1000 °C of calcinations, the as prepared amorphous alumina transformed in to crystalline state, while below this temperature, i.e. at 800 °C calcination, the powder was amorphous in nature. The point of zero charge (PZC) of the desired sample was estimated by the salt addition method.

  14. Heat Transfer in Boiling Dilute Emulsion with Strong Buoyancy

    NASA Astrophysics Data System (ADS)

    Freeburg, Eric Thomas

    Little attention has been given to the boiling of emulsions compared to that of boiling in pure liquids. The advantages of using emulsions as a heat transfer agent were first discovered in the 1970s and several interesting features have since been studied by few researchers. Early research focuses primarily on pool and flow boiling and looks to determine a mechanism by which the boiling process occurs. This thesis looks at the boiling of dilute emulsions in fluids with strong buoyant forces. The boiling of dilute emulsions presents many favorable characteristics that make it an ideal agent for heat transfer. High heat flux electronics, such as those seen in avionics equipment, produce high heat fluxes of 100 W/cm2 or more, but must be maintained at low temperatures. So far, research on single phase convection and flow boiling in small diameter channels have yet to provide an adequate solution. Emulsions allow the engineer to tailor the solution to the specific problem. The fluid can be customized to retain the high thermal conductivity and specific heat capacity of the continuous phase while enhancing the heat transfer coefficient through boiling of the dispersed phase component. Heat transfer experiments were carried out with FC-72 in water emulsions. FC-72 has a saturation temperature of 56 °C, far below that of water. The parameters were varied as follows: 0% ≤ epsilon ≤ 1% and 1.82 x 1012 ≤ RaH ≤ 4.42 x 1012. Surface temperatures along the heated surface reached temperature that were 20 °C in excess of the dispersed phase saturation temperature. An increase of ˜20% was seen in the average Nusselt numbers at the highest Rayleigh numbers. Holography was used to obtain images of individual and multiple FC-72 droplets in the boundary layer next to the heated surface. The droplet diameters ranged from 0.5 mm to 1.3 mm. The Magnus effect was observed when larger individual droplets were injected into the boundary layer, causing the droplets to be pushed

  15. Aspects of subcooled boiling

    SciTech Connect

    Bankoff, S.G.

    1997-12-31

    Subcooled boiling boiling refers to boiling from a solid surface where the bulk liquid temperature is below the saturation temperature (subcooled). Two classes are considered: (1) nucleate boiling, where, for large subcoolings, individual bubbles grow and collapse while remaining attached to the solid wall, and (2) film boiling, where a continuous vapor film separates the solid from the bulk liquid. One mechanism by which subcooled nucleate boiling results in very large surface heat transfer coefficient is thought to be latent heat transport within the bubble, resulting from simultaneous evaporation from a thin residual liquid layer at the bubble base, and condensation at the polar bubble cap. Another is the increased liquid microconvection around the oscillating bubble. Two related problems have been attacked. One is the rupture of a thin liquid film subject to attractive and repulsive dispersion forces, leading to the formation of mesoscopic drops, which then coalesce and evaporate. Another is the liquid motion in the vicinity of an oscillating contact line, where the bubble wall is idealized as a wedge of constant angle sliding on the solid wall. The subcooled film boiling problem has been attacked by deriving a general long-range nonlinear evolution equation for the local thickness of the vapor layer. Linear and weakly-nonlinear stability results have been obtained. A number of other related problems have been attacked.

  16. Spacelab Mission 3 experiment descriptions

    NASA Technical Reports Server (NTRS)

    Hill, C. K. (Editor)

    1982-01-01

    The Spacelab 3 mission is the first operational flight of Spacelab aboard the shuttle transportation system. The primary objectives of this mission are to conduct application, science, and technology experimentation that requires the low gravity environment of Earth orbit and an extended duration, stable vehicle attitude with emphasis on materials processing. This document provides descriptions of the experiments to be performed during the Spacelab 3 mission.

  17. The experiments of LIPS 3

    NASA Technical Reports Server (NTRS)

    Severns, J. G.; Hobbs, R. M.; Elliott, N. P.; Towsley, R. H.; Virshup, G. F.

    1989-01-01

    LIPS 3 is a member of the Living Plume Shield series of spacecraft. In each LIPS project, the plume shield, a simple sheet metal cone, was structurally stiffened, and an active satellite was then built around it. The original purpose of the plume shield was to prevent the plume from solid propellent engines, which are fired outside the atmosphere after the aerodynamic shroud is jettisoned, from reaching the primary payload. The surface of LIPS 3 facing the plume also functioned in this manner, but the anterior surfaces were unaffected, and it was there that all solar arrays, sensors, and experiments were mounted. The purpose of LIPS 3 was to provide a test bed for new space power sources. With the long delays projected for schedules of the STS and other major launch systems, it appeared that a decade might pass before long term flight data could be obtained on many new and innovative power sources. The fact that a launch scheduled for early in 1987 required a plume shield was seen as a unique opportunity to obtain some of this data in a timely manner. The LIPS 3 system, the experiments placed aboard, and the experiment data acquisition subsystem are described. Various problems were encountered during integration and after launch; those which appear to effect the accuracy of experimental results are discussed. A preliminary description is given of the accuracy of the flight experiment data.

  18. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  19. Effect of boiling regime on melt stream breakup in water

    SciTech Connect

    Spencer, B.W.; Gabor, J.D.; Cassulo, J.C.

    1986-01-01

    A study has been performed examining the breakup and mixing behavior of an initially coherent stream of high-density melt as it flows downward through water. This work has application to the quenching of molten core materials as they drain downward during a postulated severe reactor accident. The study has included examination of various models of breakup distances based upon interfacial instabilities dominated either by liquid-liquid contact or by liquid-vapor contact. A series of experiments was performed to provide a data base for assessment of the various modeling approaches. The experiments involved Wood's metal (T/sub m/ = 73/sup 0/C, rho = 9.2 g/cm/sup 3/, d/sub j/ = 20 mm) poured into a deep pool of water. The temperature of the water and wood's metal were varied to span the range from single-phase, liquid-liquid contact to the film boiling regime. Experiment results showed that breakup occurred largely as a result of the spreading and entrainment from the leading edge of the jet. However, for streams of sufficient lengths a breakup length could be discerned at which there was no longer a coherent central core of the jet to feed the leading edge region. The erosion of the vertical trailing column is by Kelvin-Helmoltz instabilities and related disengagement of droplets from the jet into the surrounding fluid. For conditions of liquid-liquid contact, the breakup length has been found to be about 20 jet diameters; when substantial vapor is produced at the interface due to heat transfer from the jet to the water, the breakup distance was found to range to as high as 50 jet diameters. The former values are close to the analytical prediction of Taylor, whereas the latter values are better predicted by the model of Epstein and Fauske.

  20. Outbreak of boils in an Alaskan village

    PubMed Central

    Landen, Michael G; McCumber, Barbara J; Asay, Elvin D; Egeland, Grace M

    2000-01-01

    Objective To determine whether taking steam baths was associated with furunculosis and to evaluate possible risk factors for the occurrence of boils during a large outbreak in Alaska. Design A cohort study of village residents, a case-control study, and assessment of environmental cultures taken from steam baths. Setting Village in southwestern Alaska. Participants 1 adult member from 77 of the 92 house-holds in the village was interviewed; 115 residents with at least one boil occurring between January 1 and December 12, 1996 were considered to be cases; 209 residents without a boil acted as the control group. All 459 village residents were included in the cohort study. Main outcome measure Rate of infection among all residents and residents who regularly took steam baths, risk factors for infection, and relative risk of infection. Results 115 people (25%) had had at least one boil. Men were more likely to have had a boil than women (relative risk 1.5; 95% confidence interval 1.1 to 2.2). The highest rate of infection was among people ages 25-34 years (32/76; 42%). No children younger than 2 years had had boils. Boils were associated with using a steam bath (odds ratio 8.1; 3.3 to 20.1). Among those who used a steam bath, the likelihood of developing boils was reduced by routinely sitting on a towel while bathing, which women were more likely to do, and bathing with fewer than 8 people. Of the 93 samples taken from steam baths, one Staphylococcus aureus isolate was obtained from a bench in an outer dressing room. Conclusion Using a steam bath was associated with developing boils in this outbreak in a village in Alaska. People should be advised to sit on towels while using steam baths. PMID:10778372

  1. Indication of reactor ν(e) disappearance in the Double Chooz experiment.

    PubMed

    Abe, Y; Aberle, C; Akiri, T; dos Anjos, J C; Ardellier, F; Barbosa, A F; Baxter, A; Bergevin, M; Bernstein, A; Bezerra, T J C; Bezrukhov, L; Blucher, E; Bongrand, M; Bowden, N S; Buck, C; Busenitz, J; Cabrera, A; Caden, E; Camilleri, L; Carr, R; Cerrada, M; Chang, P-J; Chimenti, P; Classen, T; Collin, A P; Conover, E; Conrad, J M; Cormon, S; Crespo-Anadón, J I; Cribier, M; Crum, K; Cucoanes, A; D'Agostino, M V; Damon, E; Dawson, J V; Dazeley, S; Dierckxsens, M; Dietrich, D; Djurcic, Z; Dracos, M; Durand, V; Efremenko, Y; Elnimr, M; Endo, Y; Etenko, A; Falk, E; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Fernandes, S M; Franco, D; Franke, A J; Franke, M; Furuta, H; Gama, R; Gil-Botella, I; Giot, L; Göger-Neff, M; Gonzalez, L F G; Goodman, M C; Goon, J T M; Greiner, D; Guillon, B; Haag, N; Hagner, C; Hara, T; Hartmann, F X; Hartnell, J; Haruna, T; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G A; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L; Kamyshkov, Y; Kaplan, D M; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Kibe, Y; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; Langbrandtner, C; Lasserre, T; Letourneau, A; Lhuillier, D; Lima, H P; Lindner, M; Liu, Y; López-Castanõ, J M; LoSecco, J M; Lubsandorzhiev, B K; Lucht, S; McKee, D; Maeda, J; Maesano, C N; Mariani, C; Maricic, J; Martino, J; Matsubara, T; Mention, G; Meregaglia, A; Miletic, T; Milincic, R; Milzstajn, A; Miyata, H; Motta, D; Mueller, Th A; Nagasaka, Y; Nakajima, K; Novella, P; Obolensky, M; Oberauer, L; Onillon, A; Osborn, A; Ostrovskiy, I; Palomares, C; Peeters, S J M; Pepe, I M; Perasso, S; Perrin, P; Pfahler, P; Porta, A; Potzel, W; Queval, R; Reichenbacher, J; Reinhold, B; Remoto, A; Reyna, D; Röhling, M; Roth, S; Rubin, H A; Sakamoto, Y; Santorelli, R; Sato, F; Schönert, S; Schoppmann, S; Schwan, U; Schwetz, T; Shaevitz, M H; Shrestha, D; Sida, J-L; Sinev, V; Skorokhvatov, M; Smith, E; Spitz, J; Stahl, A; Stancu, I; Strait, M; Stüken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Sun, Z; Svoboda, R; Tabata, H; Tamura, N; Terao, K; Tonazzo, A; Toups, M; Trinh Thi, H H; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yanovitch, E; Yermia, F; Zbiri, K; Zimmer, V

    2012-03-30

    The Double Chooz experiment presents an indication of reactor electron antineutrino disappearance consistent with neutrino oscillations. An observed-to-predicted ratio of events of 0.944±0.016(stat)±0.040(syst) was obtained in 101 days of running at the Chooz nuclear power plant in France, with two 4.25 GW(th) reactors. The results were obtained from a single 10 m(3) fiducial volume detector located 1050 m from the two reactor cores. The reactor antineutrino flux prediction used the Bugey4 flux measurement after correction for differences in core composition. The deficit can be interpreted as an indication of a nonzero value of the still unmeasured neutrino mixing parameter sin(2)2θ(13). Analyzing both the rate of the prompt positrons and their energy spectrum, we find sin(2)2θ(13)=0.086±0.041(stat)±0.030(syst), or, at 90% C.L., 0.017

  2. Characterization of the homogeneous reactor experiment No. 2 (HRE) impoundment

    SciTech Connect

    Stansfield, R.G.; Francis, C.W.

    1986-07-01

    A characterization study was conducted on a radioactive waste impoundment for the Homogeneous Reactor Experiment No. 2 at the Oak Ridge National Laboratory, to provide information necessary for its proper disposition. The impoundment received low-level radioactive wastes from 1957 until 1963. In 1970, the pond was backfilled with clay soil and the impoundment capped with asphaltic concrete, but no wastes were removed. The mixed soil fill and sediment, approximately 4.8 m deep, was sampled using soil boring methods. The samples were analyzed to determine if the material would classify as a hazardous waste under regulatory definitions promulgated in accordance with the Resource Conservation and Recovery Act (RCRA). If the soil and sediment mixture contained RCRA-defined hazardous waste, it would be subject to the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). Chemical analyses indicate that the sampled material does not contain hazardous chemical constituents above the levels permitted by RCRA regulations. The sediment was found to contain an estimated radioactivity inventory of approximately 2700 GBq (70 Ci) of /sup 90/Sr and 600 GBq (16 Ci) of /sup 137/Cs.

  3. Decommissioning of the Molten Salt Reactor Experiment: A technical evaluation

    SciTech Connect

    Notz, K.J.

    1988-01-01

    This report completes a technical evaluation of decommissioning planning for the former Molten Salt Reactor Experiment, which was shut down in December, 1969. The key issues revolve around the treatment and disposal of some five tons of solid fuel salt which contains over 30 kg of fissionable uranium-233 plus fission products and higher actinides. The chemistry of this material is complicated by the formation of elemental fluorine via a radiolysis reaction under certain conditions. Supporting studies carried out as part of this evaluation include (a) a broad scope analysis of possible options for storage/disposal of the salts, (b) calculation of nuclide decay in future years, (c) technical evaluation of the containment facility and hot cell penetrations, (d) review and update of surveillance and maintenance procedures, (e) measurements of facility groundwater radioactivity and sump pump operation, (f) laboratory studies of the radiolysis reaction, and (g) laboratory studies which resulted in finding a suitable getter for elemental fluorine. In addition, geologic and hydrologic factors of the surrounding area were considered, and also the implications of entombment of the fuel in-place with concrete. The results of this evaluation show that the fuel salt cannot be left in its present form and location permanently. On the other hand, extended storage in its present form is quite acceptable for 20 to 30 years, or even longer. For continued storage in-place, some facility modifications are recommended. 30 refs., 5 figs., 9 tabs.

  4. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  5. Removal of 3-monochlorophenol in Anaerobic Baffled Reactor

    NASA Astrophysics Data System (ADS)

    Farrokhi, M.; Mesdaghinia, A. R.

    Monochlorophenols, particularly 3-monochlorophenol are resistant to biodegradation than many poly- chlorinated compound. The presence of chlorophenols in water has become a significant pollution problem. The biodegradation of 3-monochlorophenol was investigated by Anaerobic Baffled Reactor (ABR). A mixture of 3-MCP and glucose as synthetic wastewater was treated in a laboratory scale. The ABR reactor volume was 9 L and wastewater with different concentration of 3-MCP and hydraulic retention time of 2 days, was treated. The parameters that determined include COD, volatile suspended solids (VSS) and 3-MCP concentration. The initial sludge for reactor feeding was isolated from full scale ABR that was operated in an industrial complex. When the concentration of 3-MCP was 10 mg LG1 and less, 3-MCP was able to be metabolized to mineral end products and the COD removal was about 90% and more. With increasing of 3-MCP concentration from 10 to 30 mg LG1, the degradation of 3-MCP and COD removal were decreased significantly (p< 0.001). The variations of VSS in different concentration of 3-MCP was significant (p< 0.001). Increasing of COD to 2000 mg LG1, improved the reactor performance and reduced the toxicity of 3-MCP. In the toxicity evaluation of 3-MCP on ABR biomass, the ratio of glucose/3-MCP is more important parameter than 3-MCP concentration.

  6. Consolidated fuel reprocessing program: Criticality experiments with fast test reactor fuel pins in an organic moderator

    SciTech Connect

    Bierman, S.R.

    1986-12-01

    The results obtained in a series of criticality experiments performed as part of a joint program on criticality data development between the United States Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan are presented in this report along with a complete description of the experiments. The experiments involved lattices of Fast Test Reactor (FTR) fuel pins in an organic moderator mixture similar to that used in the solvent extraction stage of fuel reprocessing. The experiments are designed to provide data for direct comparison with previously performed experimental measurements with water moderated lattices of FTR fuel pins. The same lattice arrangements and FTR fuel pin types are used in these organic moderated experimental assemblies as were used in the water moderated experiments. The organic moderator is a mixture of 38 wt % tributylphosphate in a normal paraffin hydrocarbon mixture of C{sub 11}H{sub 24} to C{sub 15}H{sub 32} molecules. Critical sizes of 1054.8, 599.2, 301.8, 199.5 and 165.3 fuel pins were obtained respectively for organic moderated lattices having 0.761 cm, 0.968 cm, 1.242 cm, 1.537 cm and 1.935 cm square lattice pitches as compared to 1046.9, 571.9, 293.9, 199.7 and 165.1 fuel pins for the same lattices water moderated.

  7. Analysis of Nuclear Reactor Background Radiation for Neutrino Experiments

    NASA Astrophysics Data System (ADS)

    Leblanc, Ricky; Blackmon, J. C.; Rasco, B. C.; Mumm, H. P.; mTC; NuLat Collaboration

    2015-10-01

    Prior measurements of reactor antineutrinos have found a lower flux than expected. Precision measurements of antineutrino energy spectra are important for understanding the anomaly, reactor safeguards, and nuclear nonproliferation. Antineutrino detector designs rely on good characterization of gamma-ray and neutron backgrounds near the reactor core. To study the gamma-ray background at the NIST research reactor, spectra were collected using a 6.25 cm diameter × 5.5 cm germanium detector. We analyzed the measured spectra using simulations of the detector response using the GEANT4 toolkit to determine background fluxes and build a background model that will be used to understand shielding requirements and the impact of backgrounds on potential short-baseline reactor antineutrino studies at NIST. This work supported by the National Science Foundation and LSU.

  8. Odd-Boiled Eggs

    ERIC Educational Resources Information Center

    Kaminsky, Kenneth; Scheman, Naomi

    2010-01-01

    At a Shabbat lunch in Madrid not long ago, the conversation turned to the question of boiling eggs. One of the guests mentioned that a Dutch rabbi he knew had heard that in order to make it more likely that boiled eggs be kosher, you should add an egg to the pot if the number you began with was even. According to the laws of Kashruth, Jews may not…

  9. Reactor ν̄e disappearance in the Double Chooz experiment

    DOE PAGESBeta

    Abe, Y.; Aberle, C.; dos Anjos, J. C.; Barriere, J. C.; Bergevin, M.; Bernstein, A.; Bezerra, T. J. C.; Bezrukhov, L.; Blucher, E.; Bowden, N. S.; et al

    2012-09-18

    The Double Chooz experiment has observed 8249 candidate electron antineutrino events in 227.93 live days with 33.71 GW-ton-years (reactor power×detector mass×live time) exposure using a 10.3 m³ fiducial volume detector located at 1050 m from the reactor cores of the Chooz nuclear power plant in France. The expectation in case of θ₁₃=0 is 8937 events. The deficit is interpreted as evidence of electron antineutrino disappearance. From a rate plus spectral shape analysis we find sin²2θ₁₃=0.109±0.030(stat)±0.025(syst). The data exclude the no-oscillation hypothesis at 99.8% CL (2.9σ).

  10. Reactor ν̄e disappearance in the Double Chooz experiment

    SciTech Connect

    Abe, Y.; Aberle, C.; dos Anjos, J. C.; Barriere, J. C.; Bergevin, M.; Bernstein, A.; Bezerra, T. J. C.; Bezrukhov, L.; Blucher, E.; Bowden, N. S.; Buck, C.; Busenitz, J.; Cabrera, A.; Caden, E.; Camilleri, L.; Carr, R.; Cerrada, M.; Chang, P.-J.; Chimenti, P.; Classen, T.; Collin, A. P.; Conover, E.; Conrad, J. M.; Crespo-Anadón, J. I.; Crum, K.; Cucoanes, A.; D’Agostino, M. V.; Damon, E.; Dawson, J. V.; Dazeley, S.; Dietrich, D.; Djurcic, Z.; Dracos, M.; Durand, V.; Ebert, J.; Efremenko, Y.; Elnimr, M.; Etenko, A.; Fallot, M.; Fechner, M.; von Feilitzsch, F.; Felde, J.; Franco, D.; Franke, A. J.; Franke, M.; Furuta, H.; Gama, R.; Gil-Botella, I.; Giot, L.; Göger-Neff, M.; Gonzalez, L. F. G.; Goodman, M. C.; Goon, J. TM.; Greiner, D.; Haag, N.; Hagner, C.; Hara, T.; Hartmann, F. X.; Haser, J.; Hatzikoutelis, A.; Hayakawa, T.; Hofmann, M.; Horton-Smith, G. A.; Hourlier, A.; Ishitsuka, M.; Jochum, J.; Jollet, C.; Jones, C. L.; Kaether, F.; Kalousis, L. N.; Kamyshkov, Y.; Kaplan, D. M.; Kawasaki, T.; Keefer, G.; Kemp, E.; de Kerret, H.; Kibe, Y.; Konno, T.; Kryn, D.; Kuze, M.; Lachenmaier, T.; Lane, C. E.; Langbrandtner, C.; Lasserre, T.; Letourneau, A.; Lhuillier, D.; Lima, H. P.; Lindner, M.; López-Castanõ, J. M.; LoSecco, J. M.; Lubsandorzhiev, B. K.; Lucht, S.; McKee, D.; Maeda, J.; Maesano, C. N.; Mariani, C.; Maricic, J.; Martino, J.; Matsubara, T.; Mention, G.; Meregaglia, A.; Miletic, T.; Milincic, R.; Miyata, H.; Mueller, Th. A.; Nagasaka, Y.; Nakajima, K.; Novella, P.; Obolensky, M.; Oberauer, L.; Onillon, A.; Osborn, A.; Ostrovskiy, I.; Palomares, C.; Pepe, I. M.; Perasso, S.; Perrin, P.; Pfahler, P.; Porta, A.; Potzel, W.; Reichenbacher, J.; Reinhold, B.; Remoto, A.; Röhling, M.; Roncin, R.; Roth, S.; Sakamoto, Y.; Santorelli, R.; Sato, F.; Schönert, S.; Schoppmann, S.; Schwetz, T.; Shaevitz, M. H.; Shimojima, S.; Shrestha, D.; Sida, J-L.; Sinev, V.; Skorokhvatov, M.; Smith, E.; Spitz, J.; Stahl, A.; Stancu, I.; Stokes, L. F. F.; Strait, M.; Stüken, A.; Suekane, F.; Sukhotin, S.; Sumiyoshi, T.; Sun, Y.; Svoboda, R.; Terao, K.; Tonazzo, A.; Toups, M.; Trinh Thi, H. H.; Valdiviesso, G.; Veyssiere, C.; Wagner, S.; Watanabe, H.; White, B.; Wiebusch, C.; Winslow, L.; Worcester, M.; Wurm, M.; Yermia, F.; Zimmer, V.

    2012-09-18

    The Double Chooz experiment has observed 8249 candidate electron antineutrino events in 227.93 live days with 33.71 GW-ton-years (reactor power×detector mass×live time) exposure using a 10.3 m³ fiducial volume detector located at 1050 m from the reactor cores of the Chooz nuclear power plant in France. The expectation in case of θ₁₃=0 is 8937 events. The deficit is interpreted as evidence of electron antineutrino disappearance. From a rate plus spectral shape analysis we find sin²2θ₁₃=0.109±0.030(stat)±0.025(syst). The data exclude the no-oscillation hypothesis at 99.8% CL (2.9σ).

  11. Controlled tissue emulsification produced by high intensity focused ultrasound shock waves and millisecond boiling

    PubMed Central

    Khokhlova, Tatiana D.; Canney, Michael S.; Khokhlova, Vera A.; Sapozhnikov, Oleg A.; Crum, Lawrence A.; Bailey, Michael R.

    2011-01-01

    In high intensity focused ultrasound (HIFU) applications, tissue may be thermally necrosed by heating, emulsified by cavitation, or, as was recently discovered, emulsified using repetitive millisecond boiling caused by shock wave heating. Here, this last approach was further investigated. Experiments were performed in transparent gels and ex vivo bovine heart tissue using 1, 2, and 3 MHz focused transducers and different pulsing schemes in which the pressure, duty factor, and pulse duration were varied. A previously developed derating procedure to determine in situ shock amplitudes and the time-to-boil was refined. Treatments were monitored using B-mode ultrasound. Both inertial cavitation and boiling were observed during exposures, but emulsification occurred only when shocks and boiling were present. Emulsified lesions without thermal denaturation were produced with shock amplitudes sufficient to induce boiling in less than 20 ms, duty factors of less than 0.02, and pulse lengths shorter than 30 ms. Higher duty factors or longer pulses produced varying degrees of thermal denaturation combined with mechanical emulsification. Larger lesions were obtained using lower ultrasound frequencies. The results show that shock wave heating and millisecond boiling is an effective and reliable way to emulsify tissue while monitoring the treatment with ultrasound. PMID:22088025

  12. Controlled tissue emulsification produced by high intensity focused ultrasound shock waves and millisecond boiling.

    PubMed

    Khokhlova, Tatiana D; Canney, Michael S; Khokhlova, Vera A; Sapozhnikov, Oleg A; Crum, Lawrence A; Bailey, Michael R

    2011-11-01

    In high intensity focused ultrasound (HIFU) applications, tissue may be thermally necrosed by heating, emulsified by cavitation, or, as was recently discovered, emulsified using repetitive millisecond boiling caused by shock wave heating. Here, this last approach was further investigated. Experiments were performed in transparent gels and ex vivo bovine heart tissue using 1, 2, and 3 MHz focused transducers and different pulsing schemes in which the pressure, duty factor, and pulse duration were varied. A previously developed derating procedure to determine in situ shock amplitudes and the time-to-boil was refined. Treatments were monitored using B-mode ultrasound. Both inertial cavitation and boiling were observed during exposures, but emulsification occurred only when shocks and boiling were present. Emulsified lesions without thermal denaturation were produced with shock amplitudes sufficient to induce boiling in less than 20 ms, duty factors of less than 0.02, and pulse lengths shorter than 30 ms. Higher duty factors or longer pulses produced varying degrees of thermal denaturation combined with mechanical emulsification. Larger lesions were obtained using lower ultrasound frequencies. The results show that shock wave heating and millisecond boiling is an effective and reliable way to emulsify tissue while monitoring the treatment with ultrasound. PMID:22088025

  13. Validation of HELIOS Neutron Cross-Section Library for RBMK Reactors Against the Data From the Critical Facility Experiments

    SciTech Connect

    Jasiulevicius, Audrius; Sehgal, Bal Raj

    2002-07-01

    The RBMK reactors are channel type, water-cooled and graphite moderated reactors. The first RBMK type electricity production reactor was put on-line in 1973. Currently there are 13 operating reactors of this type. Two of the RBMK-1500 reactors are at the Ignalina NPP in Lithuania. Experimental Critical Facility for RBMK reactors, located at Kurchatov Institute, Moscow was designed to carry out critical reactivity experiments on assemblies, which imitate parts of the RBMK reactor core. The facility is composed of Control and Protection Rods (CPR's), fuel assemblies with different enrichment in U-235 and other elements, typical for RBMK reactor core loadings, e.g. additional absorber assemblies, CPR imitators, etc. A simulation of a set of the experiments, performed at the Experimental Critical Facility, was carried out at the Royal Institute of Technology (RIT), Nuclear Power Safety Division, using CORETRAN 3-D neutron dynamics code. The neutron cross sections for assemblies were calculated using HELIOS code. The aim of this work was to evaluate capabilities of the HELIOS code to provide correct cross section data for the RBMK reactor. The calculation results were compared to the similar CORETRAN calculations, when employing WIMS-D4 code generated cross section data. For some of the experiments, where calculation results with CASMO-4 code generated cross sections are available, the comparison is also performed against CASMO-4 results. Eleven different experiments were simulated. Experiments differ in size of the facility core (number of assemblies loaded): from simple core loadings, composed only of a few fuel assemblies, to complicated configurations, which represent a part of the RBMK reactor core. Diverse types of measurements were carried out during these experiments: reactivity, neutron flux distributions (both axial and radial), rod reactivity worth and the voiding effects. Results of the reactivity measurements and relative neutron flux distributions were

  14. Heavy reflector experiments in the IPEN/MB-01 reactor: Stainless steel, carbon steel and nickel

    SciTech Connect

    Santos, Adimir dos; Andrade e Silva, Graciete Simoes de; Jerez, Rogerio; Liambos Mura, Luis Felipe; Fuga, Rinaldo

    2013-05-06

    New experiments devoted to the measurements of physical parameters of a light water core surrounded by a heavy reflector were performed in the IPEN/MB-01 research reactor facility. These experiments comprise three sets of heavy reflector (SS-304, Carbon Steel, and Nickel) in a form of laminates around 3 mm thick. Each set was introduced individually in the west face of the core of the IPEN/MB-01 reactor. The aim here is to provide high quality experimental data for the interpretation and validation of the SS-304 heavy reflector calculation methods. The experiments of Carbon Steel, which is composed mainly of iron, and Nickel were performed to provide a consistent and an interpretative check for the SS-304 reflector experiment. The experimental results comprise critical control bank positions, temperatures and reactivities as a function of the number of the plates. Particularly to the case of Nickel, the experimental data are unique of its kind. The theoretical analysis was performed by MCNP-5 with the nuclear data library ENDF/B-VII.0. It was shown that this nuclear data library has a very good performance up to thirteen plates and overestimates the reactivity for higher number of plates independently of the type of the reflector.

  15. Measuring of fissile isotope partial antineutrino spectra in direct experiment at nuclear reactor

    SciTech Connect

    Sinev, V. V.

    2009-11-15

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta-decay reaction positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

  16. CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling

    SciTech Connect

    Fan-Bill Cheung; Joy L. Rempe

    2004-06-01

    In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of the reactor cavity during a severe accident. As part of a joint Korean – United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the SBLB (Subscale Boundary Layer Boiling) facility at Penn State using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.

  17. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2006-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  18. Effect of rolling motion on critical heat flux for subcooled flow boiling in vertical tube

    SciTech Connect

    Hwang, J. S.; Park, I. U.; Park, M. Y.; Park, G. C.

    2012-07-01

    This paper presents defining characteristics of the critical heat flux (CHF) for the boiling of R-134a in vertical tube operation under rolling motion in marine reactor. It is important to predict CHF of marine reactor having the rolling motion in order to increase the safety of the reactor. Marine Reactor Moving Simulator (MARMS) tests are conducted to measure the critical heat flux using R-134a flowing upward in a uniformly heated vertical tube under rolling motion. MARMS was rotated by motor and mechanical power transmission gear. The CHF tests were performed in a 9.5 mm I.D. test section with heated length of 1 m. Mass fluxes range from 285 to 1300 kg m{sup -2}s{sup -1}, inlet subcooling from 3 to 38 deg. C and outlet pressures from 13 to 24 bar. Amplitudes of rolling range from 15 to 40 degrees and periods from 6 to 12 sec. To convert the test conditions of CHF test using R-134a in water, Katto's fluid-to-fluid modeling was used in present investigation. A CHF correlation is presented which accounts for the effects of pressure, mass flux, inlet subcooling and rolling angle over all conditions tested. Unlike existing transient CHF experiments, CHF ratio of certain mass flux and pressure are different in rolling motion. For the mass fluxes below 500 kg m{sup -2}s{sup -1} at 13, 16 (region of relative low mass flux), CHF ratio was decreased but was increased above that mass flux (region of relative high mass flux). Moreover, CHF tend to enhance in entire mass flux at 24 bar. (authors)

  19. Boiling radial flow in fractures of varying wall porosity

    SciTech Connect

    Barnitt, Robb Allan

    2000-06-01

    The focus of this report is the coupling of conductive heat transfer and boiling convective heat transfer, with boiling flow in a rock fracture. A series of experiments observed differences in boiling regimes and behavior, and attempted to quantify a boiling convection coefficient. The experimental study involved boiling radial flow in a simulated fracture, bounded by a variety of materials. Nonporous and impermeable aluminum, highly porous and permeable Berea sandstone, and minimally porous and permeable graywacke from The Geysers geothermal field. On nonporous surfaces, the heat flux was not strongly coupled to injection rate into the fracture. However, for porous surfaces, heat flux, and associated values of excess temperature and a boiling convection coefficient exhibited variation with injection rate. Nucleation was shown to occur not upon the visible surface of porous materials, but a distance below the surface, within the matrix. The depth of boiling was a function of injection rate, thermal power supplied to the fracture, and the porosity and permeability of the rock. Although matrix boiling beyond fracture wall may apply only to a finite radius around the point of injection, higher values of heat flux and a boiling convection coefficient may be realized with boiling in a porous, rather than nonporous surface bounded fracture.

  20. Multiphysics modeling of two-phase film boiling within porous corrosion deposits

    NASA Astrophysics Data System (ADS)

    Jin, Miaomiao; Short, Michael

    2016-07-01

    Porous corrosion deposits on nuclear fuel cladding, known as CRUD, can cause multiple operational problems in light water reactors (LWRs). CRUD can cause accelerated corrosion of the fuel cladding, increase radiation fields and hence greater exposure risk to plant workers once activated, and induce a downward axial power shift causing an imbalance in core power distribution. In order to facilitate a better understanding of CRUD's effects, such as localized high cladding surface temperatures related to accelerated corrosion rates, we describe an improved, fully-coupled, multiphysics model to simulate heat transfer, chemical reactions and transport, and two-phase fluid flow within these deposits. Our new model features a reformed assumption of 2D, two-phase film boiling within the CRUD, correcting earlier models' assumptions of single-phase coolant flow with wick boiling under high heat fluxes. This model helps to better explain observed experimental values of the effective CRUD thermal conductivity. Finally, we propose a more complete set of boiling regimes, or a more detailed mechanism, to explain recent CRUD deposition experiments by suggesting the new concept of double dryout specifically in thick porous media with boiling chimneys.

  1. Electrohydrodynamic Pool Boiling in Reduced Gravity

    NASA Technical Reports Server (NTRS)

    Shaw, Benjamin D.; Stahl, S. L.

    1996-01-01

    This research is concerned with studying the effects of applied electric fields on pool boiling in a reduced-gravity environment. Experiments are conducted at the NASA Lewis 2.2 sec Drop tower using a drop rig constructed at UC Davis. In the experiments, a platinum wire is heated while immersed in saturated liquid refrigerants (FC-72 and FC-87), or water, causing vapor formation at the wire surface. Electric fields are applied between the wire surface and an outer screen electrode that surrounds the wire. Preliminary normal-gravity experiments with water have demonstrated that applied electric fields generated by the rig electronics can influence boiling characteristics. Reduced-gravity experiments will be performed in the summer of 1996. The experiments will provide fundamental data on electric field strengths required to disrupt film boiling (for various wire heat generation input rates) in reduced gravity for a cylindrical geometry. The experiments should also shed light on the roles of characteristic bubble generation times and charge relaxation times in determining the effects of electric fields on pool boiling. Normal-gravity comparison experiments will also be performed.

  2. Leukemia in the proximity of a German boiling-water nuclear reactor: evidence of population exposure by chromosome studies and environmental radioactivity.

    PubMed

    Schmitz-Feuerhake, I; Dannheim, B; Heimers, A; Oberheitmann, B; Schröder, H; Ziggel, H

    1997-12-01

    Exceptional elevation of children's leukemia appearing 5 years after the 1983 startup of the Krümmel nuclear power plant, accompanied by a significant increase of adult leukemia cases, led to investigations of radiation exposures of the population living near the plant. The rate of dicentric chromosomes in peripheral lymphocytes of seven parents of children with leukemia and in 14 other inhabitants near the plant was significantly elevated and indicated ongoing exposures over the years of its operation. These findings led to the hypothesis that chronic reactor leakages had occurred. This assumption is support by identification of artificial radioactivity in air, rainwater, soil and vegetation by the environmental monitoring program at the nuclear power plant. Calculations of the corresponding source terms show that emissions must have been well above authorized annual limits. Bone marrow doses supposedly result primarily through incorporation of bone-seeking beta- and alpha-emitters. PMID:9467072

  3. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  4. Radiolysis of boiling water

    NASA Astrophysics Data System (ADS)

    Yang, Shuang; Katsumura, Yosuke; Yamashita, Shinichi; Matsuura, Chihiro; Hiroishi, Daisuke; Lertnaisat, Phantira; Taguchi, Mitsumasa

    2016-06-01

    γ-radiolysis of boiling water has been investigated. The G-value of H2 evolution was found to be very sensitive to the purity of water. In high-purity water, both H2 and O2 gases were formed in the stoichiometric ratio of 2:1; a negligible amount of H2O2 remained in the liquid phase. The G-values of H2 and O2 gas evolution depend on the dose rate: lower dose rates produce larger yields. To clarify the importance of the interface between liquid and gas phase for gas evolution, the gas evolution under Ar gas bubbling was measured. A large amount of H2 was detected, similar to the radiolysis of boiling water. The evolution of gas was enhanced in a 0.5 M NaCl aqueous solution. Deterministic chemical kinetics simulation elucidated the mechanism of radiolysis in boiling water.

  5. Nonstandard interaction effects on neutrino parameters at medium-baseline reactor antineutrino experiments

    NASA Astrophysics Data System (ADS)

    Ohlsson, Tommy; Zhang, He; Zhou, Shun

    2014-01-01

    Precision measurements of leptonic mixing parameters and the determination of the neutrino mass hierarchy are the primary goals of the forthcoming medium-baseline reactor antineutrino experiments, such as JUNO and RENO-50. In this work, we investigate the impact of nonstandard neutrino interactions (NSIs) on the measurements of {sin2 θ12,Δm212} and {sin2 θ13,Δm312}, and on the sensitivity to the neutrino mass hierarchy, at the medium-baseline reactor experiments by assuming a typical experimental setup. It turns out that the true mixing parameter sin2 θ12 can be excluded at a more than 3σ level if the NSI parameter ɛ or ɛ is as large as 2% in the most optimistic case. However, the discovery reach of NSI effects has been found to be small, and depends crucially on the CP-violating phases. Finally, we show that NSI effects could enhance or reduce the discrimination power of the JUNO and RENO-50 experiments between the normal and inverted neutrino mass hierarchies.

  6. Observation of Energy and Baseline Dependent Reactor Antineutrino Disappearance in the RENO Experiment.

    PubMed

    Choi, J H; Choi, W Q; Choi, Y; Jang, H I; Jang, J S; Jeon, E J; Joo, K K; Kim, B R; Kim, H S; Kim, J Y; Kim, S B; Kim, S Y; Kim, W; Kim, Y D; Ko, Y; Lee, D H; Lim, I T; Pac, M Y; Park, I G; Park, J S; Park, R G; Seo, H; Seo, S H; Seon, Y G; Shin, C D; Siyeon, K; Yang, J H; Yeo, I S; Yu, I

    2016-05-27

    The RENO experiment has analyzed about 500 live days of data to observe an energy dependent disappearance of reactor ν[over ¯]_{e} by comparing their prompt signal spectra measured in two identical near and far detectors. In the period between August of 2011 and January of 2013, the far (near) detector observed 31 541 (290 775) electron antineutrino candidate events with a background fraction of 4.9% (2.8%). The measured prompt spectra show an excess of reactor ν[over ¯]_{e} around 5 MeV relative to the prediction from a most commonly used model. A clear energy and baseline dependent disappearance of reactor ν[over ¯]_{e} is observed in the deficit of the observed number of ν[over ¯]_{e}. Based on the measured far-to-near ratio of prompt spectra, we obtain sin^{2}2θ_{13}=0.082±0.009(stat)±0.006(syst) and |Δm_{ee}^{2}|=[2.62_{-0.23}^{+0.21}(stat)_{-0.13}^{+0.12}(syst)]×10^{-3}  eV^{2}. PMID:27284648

  7. Observation of Energy and Baseline Dependent Reactor Antineutrino Disappearance in the RENO Experiment

    NASA Astrophysics Data System (ADS)

    Choi, J. H.; Choi, W. Q.; Choi, Y.; Jang, H. I.; Jang, J. S.; Jeon, E. J.; Joo, K. K.; Kim, B. R.; Kim, H. S.; Kim, J. Y.; Kim, S. B.; Kim, S. Y.; Kim, W.; Kim, Y. D.; Ko, Y.; Lee, D. H.; Lim, I. T.; Pac, M. Y.; Park, I. G.; Park, J. S.; Park, R. G.; Seo, H.; Seo, S. H.; Seon, Y. G.; Shin, C. D.; Siyeon, K.; Yang, J. H.; Yeo, I. S.; Yu, I.; RENO Collaboration

    2016-05-01

    The RENO experiment has analyzed about 500 live days of data to observe an energy dependent disappearance of reactor ν¯e by comparing their prompt signal spectra measured in two identical near and far detectors. In the period between August of 2011 and January of 2013, the far (near) detector observed 31 541 (290 775) electron antineutrino candidate events with a background fraction of 4.9% (2.8%). The measured prompt spectra show an excess of reactor ν¯e around 5 MeV relative to the prediction from a most commonly used model. A clear energy and baseline dependent disappearance of reactor ν¯e is observed in the deficit of the observed number of ν¯e. Based on the measured far-to-near ratio of prompt spectra, we obtain sin22 θ13=0.082 ±0.009 (stat)±0.006 (syst) and |Δ mee 2|=[2.6 2-0.23+0.21(stat)-0.13+0.12(syst)]×10-3 eV2 .

  8. Complete Numerical Simulation of Subcooled Flow Boiling in the Presence of Thermal and Chemical Interactions

    SciTech Connect

    V.K. Dhir

    2003-04-28

    At present, guidelines for fuel cycle designs to prevent axial offset anomalies (AOA) in pressurized water reactor (PWR) cores are based on empirical data from several operating reactors. Although the guidelines provide an ad-hoc solution to the problem, a unified approach based on simultaneous modeling of thermal-hydraulics, chemical, and nuclear interactions with vapor generation at the fuel cladding surface does not exist. As a result, the fuel designs are overly constrained with a resulting economic penalty. The objective of present project is to develop a numerical simulation model supported by laboratory experiments that can be used for fuel cycle design with respect to thermal duty of the fuel to avoid economic penalty, as well as, AOA. At first, two-dimensional numerical simulation of the growth and departure of a bubble in pool boiling with chemical interaction is considered. A finite difference scheme is used to solve the equations governing conservation of mass, momentum, energy, and species concentration. The Level Set method is used to capture the evolving liquid-vapor interface. A dilute aqueous boron solution is considered in the simulation. From numerical simulations, the dynamic change in concentration distribution of boron during the bubble growth shows that the precipitation of boron can occur near the advancing and receding liquid-vapor interface when the ambient boron concentration level is 3,000 ppm by weight. Secondly, a complete three-dimensional numerical simulation of inception, growth and departure of a single bubble subjected to forced flow parallel to the heater surface was developed. Experiments on a flat plate heater with water and with boron dissolved in the water were carried out. The heater was made out of well-polished silicon wafer. Numbers of nucleation sites and their locations were well controlled. Bubble dynamics in great details on an isolated nucleation site were obtained while varying the wall superheat, liquid subcooling

  9. Reversed boiling curve phenomenon on surfaces with interlaced wettability

    NASA Astrophysics Data System (ADS)

    Hsu, C. C.; Chiu, W. C.; Kuo, L. S.; Chen, P. H.

    2014-10-01

    We experimentally investigated the effects of contact angle difference of heterogeneous wettability surfaces on pool boiling. For surfaces exhibiting heterogeneous wettability, this study determined that the pool boiling curve experiences a superheat decrease in some regions before the system achieves the critical heat flux. In addition, oscillation of the vapor column and bubble transverse motions on the heterogeneous wettability coated surface were observed.

  10. Effects of water in film boiling over liquid metal melts

    SciTech Connect

    Greene, G.A.; Finfrock, C.; Burson, S.B.

    1986-01-01

    Liquid-liquid boiling experiments have been performed with H/sub 2/O and liquid metal melts in the 100-series test matrix (Runs 121, 126, 127) and the VE test matrix. Some of the pre-explosion unstable film boiling data as well as observations from the explosive series have been previously reported.

  11. Flooding Experiments and Modeling for Improved Reactor Safety

    SciTech Connect

    Solmos, M., Hogan, K.J., VIerow, K.

    2008-09-14

    Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.

  12. Large-scale Generation of Patterned Bubble Arrays on Printed Bi-functional Boiling Surfaces

    NASA Astrophysics Data System (ADS)

    Choi, Chang-Ho; David, Michele; Gao, Zhongwei; Chang, Alvin; Allen, Marshall; Wang, Hailei; Chang, Chih-Hung

    2016-04-01

    Bubble nucleation control, growth and departure dynamics is important in understanding boiling phenomena and enhancing nucleate boiling heat transfer performance. We report a novel bi-functional heterogeneous surface structure that is capable of tuning bubble nucleation, growth and departure dynamics. For the fabrication of the surface, hydrophobic polymer dot arrays are first printed on a substrate, followed by hydrophilic ZnO nanostructure deposition via microreactor-assisted nanomaterial deposition (MAND) processing. Wettability contrast between the hydrophobic polymer dot arrays and aqueous ZnO solution allows for the fabrication of heterogeneous surfaces with distinct wettability regions. Heterogeneous surfaces with various configurations were fabricated and their bubble dynamics were examined at elevated heat flux, revealing various nucleate boiling phenomena. In particular, aligned and patterned bubbles with a tunable departure frequency and diameter were demonstrated in a boiling experiment for the first time. Taking advantage of our fabrication method, a 6 inch wafer size heterogeneous surface was prepared. Pool boiling experiments were also performed to demonstrate a heat flux enhancement up to 3X at the same surface superheat using bi-functional surfaces, compared to a bare stainless steel surface.

  13. Large-scale Generation of Patterned Bubble Arrays on Printed Bi-functional Boiling Surfaces.

    PubMed

    Choi, Chang-Ho; David, Michele; Gao, Zhongwei; Chang, Alvin; Allen, Marshall; Wang, Hailei; Chang, Chih-hung

    2016-01-01

    Bubble nucleation control, growth and departure dynamics is important in understanding boiling phenomena and enhancing nucleate boiling heat transfer performance. We report a novel bi-functional heterogeneous surface structure that is capable of tuning bubble nucleation, growth and departure dynamics. For the fabrication of the surface, hydrophobic polymer dot arrays are first printed on a substrate, followed by hydrophilic ZnO nanostructure deposition via microreactor-assisted nanomaterial deposition (MAND) processing. Wettability contrast between the hydrophobic polymer dot arrays and aqueous ZnO solution allows for the fabrication of heterogeneous surfaces with distinct wettability regions. Heterogeneous surfaces with various configurations were fabricated and their bubble dynamics were examined at elevated heat flux, revealing various nucleate boiling phenomena. In particular, aligned and patterned bubbles with a tunable departure frequency and diameter were demonstrated in a boiling experiment for the first time. Taking advantage of our fabrication method, a 6 inch wafer size heterogeneous surface was prepared. Pool boiling experiments were also performed to demonstrate a heat flux enhancement up to 3X at the same surface superheat using bi-functional surfaces, compared to a bare stainless steel surface. PMID:27034255

  14. Large-scale Generation of Patterned Bubble Arrays on Printed Bi-functional Boiling Surfaces

    PubMed Central

    Choi, Chang-Ho; David, Michele; Gao, Zhongwei; Chang, Alvin; Allen, Marshall; Wang, Hailei; Chang, Chih-hung

    2016-01-01

    Bubble nucleation control, growth and departure dynamics is important in understanding boiling phenomena and enhancing nucleate boiling heat transfer performance. We report a novel bi-functional heterogeneous surface structure that is capable of tuning bubble nucleation, growth and departure dynamics. For the fabrication of the surface, hydrophobic polymer dot arrays are first printed on a substrate, followed by hydrophilic ZnO nanostructure deposition via microreactor-assisted nanomaterial deposition (MAND) processing. Wettability contrast between the hydrophobic polymer dot arrays and aqueous ZnO solution allows for the fabrication of heterogeneous surfaces with distinct wettability regions. Heterogeneous surfaces with various configurations were fabricated and their bubble dynamics were examined at elevated heat flux, revealing various nucleate boiling phenomena. In particular, aligned and patterned bubbles with a tunable departure frequency and diameter were demonstrated in a boiling experiment for the first time. Taking advantage of our fabrication method, a 6 inch wafer size heterogeneous surface was prepared. Pool boiling experiments were also performed to demonstrate a heat flux enhancement up to 3X at the same surface superheat using bi-functional surfaces, compared to a bare stainless steel surface. PMID:27034255

  15. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    SciTech Connect

    Farnum, E.; Scharborough, K.; Shikama, Tatsuo

    1995-04-01

    The objective of this experiment is to determine the extent of degradation during neutron irradiation of electrical and optical properties of candidate dielectric materials. The goals are to identify promising dielectrics for ITER and other fusion machines for diagnostic applications and establish the basis for optimization of candidate materials. An experiment to measure radiation-induced electrical degradation (REID) in sapphire and MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260 {degree}C to a fluence of 3{times}10{sup 24} n/m{sup 2} (E>1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m.

  16. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  17. Constraint on neutrino decay with medium-baseline reactor neutrino oscillation experiments

    NASA Astrophysics Data System (ADS)

    Abrahão, Thamys; Minakata, Hisakazu; Nunokawa, Hiroshi; Quiroga, Alexander A.

    2015-11-01

    The experimental bound on lifetime of ν 3, the neutrino mass eigenstate with the smallest ν e component, is much weaker than those of ν 1 and ν 2 by many orders of magnitude to which the astrophysical constraints apply. We argue that the future reactor neutrino oscillation experiments with medium-baseline (˜50 km), such as JUNO or RENO-50, has the best chance of placing the most stringent constraint on ν3 lifetime among all neutrino experiments which utilize the artificial source neutrinos. Assuming decay into invisible states, we show by a detailed χ 2 analysis that the ν 3 lifetime divided by its mass, τ 3 /m 3, can be constrained to be τ 3 /m 3 > 7 .5 (5 .5) × 10-11 s/eV at 95% (99%) C.L. by 100 kt·years exposure by JUNO. It may be further improved to the level comparable to the atmospheric neutrino bound by its longer run. We also discuss to what extent ν 3 decay affects mass-ordering determination and precision measurements of the mixing parameters.

  18. Lead-bismuth eutectic as advanced reactor collant : operational experience

    SciTech Connect

    Woloshun, K. A.; Watts, V.; Li, N.

    2004-01-01

    Some proposed advanced reactor concepts would be cooled by lead or lead-bismuth eutectic (LBE). An LBE test loop was designed and built at Los Alamos to develop the engineering and materials technology necessary to successfully implement LBE as a coolant (Fig. 1). Operational since December 2001, this test loop has been used to develop and demonstrate safe operation, oxygen concentration and metal corrosion control, instrumentation, thermal-hydraulic performance of heat exchangers and recuperators, and free convection and forced pumping. This paper discusses the technology development and lessons learned from the operation of this facility. A LBE test loop has been operational since December 2001. Using procedures, training, and engineering controls, this loop has operated without an accident. Continuous improvements in operation procedures and instrumentation over these years have resulted in a facility of high reliability, providing the groundwork for the use of LBE as a reactor coolant for temperatures up to 550 C.

  19. High heat flux burnout in subcooled flow boiling

    NASA Astrophysics Data System (ADS)

    Celata, G. P.; Cumo, M.; Mariani, A.

    1995-09-01

    The paper reports the results of an experimental research carried out at the Heat Transfer Division of the Energy Department, C.R. Casaccia, on the thermal hydraulic characterization of subcooled flow boiling CHF under typical conditions of thermonuclear fusion reactors, i.e. high liquid velocity and subcooling. The experiment was carried out exploring the following parameters: channel diameter (from 2.5 to 8.0 mm), heated length (10 and 15 cm), liquid velocity (from 2 to 40 m/s), exit pressure (from atmospheric to 5.0 MPa), inlet temperature (from 30 to 80 °C), channel orientation (vertical and horizontal). A maximum CHF value of 60.6 MW/m2 has been obtained under the following conditions: T in=30°, p=2.5 MPa, u=40 m/s, D=2.5 mm (smooth channel) Turbulence promoters (helically coiled wires) have been employed to further enhance the CHF attainable with subcooled flow boiling. Helically coiled wires allow an increase of 50% of the maximum CHF obtained with smooth channels.

  20. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    NASA Astrophysics Data System (ADS)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

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

    SciTech Connect

    Dr. Ronald D. Boyd

    2000-07-01

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

  2. A flow reactor study of neopentane oxidation at 8 atmospheres: Experiments and modeling

    SciTech Connect

    Wang, S.; Miller, D.L.; Cernansky, N.P.; Curran, H.J.; Pitz, W.J.; Westbrook, C.K.

    1999-08-01

    An existing detailed chemical kinetic reaction mechanism for neopentane oxidation is applied to new experimental measurements, taken in a flow reactor operating at a pressure of 8 atm. The reactor temperature ranged from 620 K to 810 K and flow rates of the reactant gases neopentane, oxygen, and nitrogen were 0.285, 7.6, and 137.1 standard liter per minute (SLM), respectively, producing an equivalence ratio of 0.3. Initial simulations identified some deficiencies in the existing model and the paper presents modifications which included upgrading the thermodynamic parameters of alkyl radical and alkylperoxy radical species, adding an alternative isomerization reaction of hydroperoxy-neopentyl-peroxy, and a multistep reaction sequence for 2-methylpropan-2-yl radical with molecular oxygen. These changes improved the calculation for the overall reactivity and the concentration profiles of the following primary products: formaldehyde, acetone, isobutene; 3,3-dimethyloxetane, methacrolein, carbon monoxide, carbon dioxide, and water. Experiments indicate that neopentane shows negative temperature coefficient behavior similar to other alkanes, though it is not as pronounced as that shown by n-pentane for example. Modeling results indicate that this behavior is caused by the {beta}-scission of the neopentyl radical and the chain propagation reactions of the hydroperoxyl-neopentyl radical.

  3. D-He-3 spherical torus fusion reactor system study

    NASA Astrophysics Data System (ADS)

    Macon, William A., Jr.

    1992-04-01

    This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

  4. Results from Boiling Temperature Measurements for Saturated Solutions in the Systems NaCl + KNO{sub 3} + H{sub 2}O, NaNO{sub 3} + KNO{sub 3} + H{sub 2}O, and NaCl + NaNO{sub 3} + KNO{sub 3} + H{sub 2}O

    SciTech Connect

    Rard, J A

    2004-10-04

    Boiling temperature measurements have been made for saturated ternary solutions of NaCl + KNO{sub 3} + H{sub 2}O and NaNO{sub 3} + KNO{sub 3} + H{sub 2}O over the full solute mole fraction range, along with the limiting binary solutions NaCl + H{sub 2}O, NaNO{sub 3} + H{sub 2}O, and KNO{sub 3} + H{sub 2}O. Boiling temperatures have also been measured for the quaternary NaCl + NaNO{sub 3} + KNO{sub 3} + H{sub 2}O mixtures with KNO{sub 3}:NaNO{sub 3} mole ratios of 1.01 and 1.19, which corresponding to the eutectic ratio and a near-eutectic ratio for the NaNO{sub 3} + KNO{sub 3} + H{sub 2}O subsystem. The maximum boiling temperature found for the NaCl + KNO{sub 3} + H{sub 2}O system is 134 C and for the NaNO{sub 3} + KNO{sub 3} + H{sub 2}O system is 160 C, but boiling temperatures as high as 196 C were measured the NaCl + NaNO{sub 3} + KNO{sub 3} + H{sub 2}O system. These mixture compositions correspond to the major mineral assemblages that are predicted to control the deliquescence relative humidity of salts found by leaching dust samples from the proposed nuclear repository at Yucca Mountain, Nevada.

  5. Boiling on horizontal surfaces coated with porous metal wicks

    SciTech Connect

    Abou-Zyan, H.Z.; Plumb, O.A.

    1997-07-01

    Boiling experiments intended to simulate heat pipe operating conditions were conducted on a copper surface covered with copper foametal and nickel fiber wicks 3.175 and 4.760 mm thick. The experiments were conducted on a horizontal surface open to the atmosphere with water as the working fluid. The experimental surface was operated like a heat pipe with distilled water supplied upstream of the heated section and transported by capillary action across a section which was adiabatic to the heated section where boiling took place. At low excess temperature, less than 10 to 20 C, the heat flux from the porous coated surfaces is comparable to or greater than that predicted for a smooth surface using the Rohsenow correlation. At higher excess temperatures corresponding to heat fluxes between 10{sup 5} and 10{sup 6} W/m{sup 2} the increase in heat flux with excess temperature is much less than that predicted by the Rohsenow correlation. When the wicks were vented by cutting slots covering 10 to 20% of the total surface area the heat flux increased, in some cases by a factor of three, for a given excess temperature. The heat flux at which the slope of the boiling curve decreased also increased for the vented surfaces. This is attributed to the provision of a low resistance path for the steam to escape providing a surface that is more highly wetted. A mathematical model for the transport with boiling in the porous wick is developed in an attempt to gain further understanding of the processes involved. The model predicts dryout conditions that are in reasonable agreement with experimental observations. However, the model predicts decreasing vapor pressure, and hence temperature, adjacent to the heated surface with increasing heat flux as a result of the decrease in relative permeability of the partially saturated wick.

  6. AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

    SciTech Connect

    Laurence Hull

    2014-10-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC 3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC 3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. All thermocouples (TCs) functioned throughout the AGC 3 experiment. There was one interval between December 18, 2012, and December 20, 2012, where 10 NULL values were reported for various TCs. These NULL values were deleted from the Nuclear Data Management and Analysis System database. All temperature data are Qualified for use by the VHTR TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the VHTR TDO Program. Total gas flow was approximately 50 sccm through the AGC 3 experiment capsule. Helium gas flow was briefly increased to 100 sccm during ATR shutdowns. At the start of the AGC 3 experiment, moisture in the outflow gas line was stuck at a constant value of 335.6174 ppmv for the first cycle (Cycle 152B). When the AGC 3 experiment capsule was reinstalled in ATR for Cycle 154B, a new moisture filter was installed. Moisture data from Cycle 152B are Failed. All moisture data from the final three cycles (Cycles 154B, 155A, and 155B) are Qualified for use by the VHTR TDO Program.

  7. Introduction to D-He(3) fusion reactors

    NASA Technical Reports Server (NTRS)

    Vlases, G. C.; Steinhauer, L. C.

    1989-01-01

    A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

  8. Film boiling of R-11 on liquid metal surfaces

    SciTech Connect

    Greene, G.A.; Irvine, T.F. Jr.

    1986-01-01

    An interesting problem is the effect of an immiscible liquid heating surface on the process of film boiling. Such surfaces raise questions concerning interface stability to disturbances, effects of gas bubbling, and vapor explosions in layered systems. The specific motivation for this study was to investigate film boiling from a liquid surface with application to cooling of molten reactor core debris by an overlying pool of reactor coolant. To investigate this phenomenon, and apparatus consisting of a nominal six-inch diameter steel vessel to hold the liquid metal and boiling fluid was constructed; coolant reservoirs, heaters, controllers, and allied instrumentation were attached. A transient energy balance was performed on the liquid metal pool by a submerged assembly of microthermocouples in the liquid metal and an array of thermocouples on the wall of the test vessel. The thermocouple data were used to determine the boiling heat flux as well as the boiling superheat. On an average basis, the deviation between the prediction of the Berenson model and the experimental data was less than one percent when Berenson was corrected for thermal radiation effects. Evidence from visualization tests of R-11 in film boiling over molten metal pools to superheats in excess of 600 K supports this conclusion. 13 refs.

  9. Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

    SciTech Connect

    Douglas W. Akers; Edwin A. Harvego

    2012-08-01

    This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and data on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.

  10. Hydraulic characterization of an activated sludge reactor with recycling system by tracer experiment and analytical models.

    PubMed

    Sánchez, F; Viedma, A; Kaiser, A S

    2016-09-15

    Fluid dynamic behaviour plays an important role in wastewater treatment. An efficient treatment requires the inexistence of certain hydraulic problems such as dead zones or short-circuiting flows. Residence time distribution (RTD) analysis is an excellent technique for detecting these inefficiencies. However, many wastewater treatment installations include water or sludge recycling systems, which prevent us from carrying out a conventional tracer pulse experiment to obtain the RTD curve of the installation. This paper develops an RTD analysis of an activated sludge reactor with recycling system. A tracer experiment in the reactor is carried out. Three analytical models, derived from the conventional pulse model, are proposed to obtain the RTD curve of the reactor. An analysis of the results is made, studying which model is the most suitable for each situation. This paper is useful to analyse the hydraulic efficiency of reactors with recycling systems. PMID:27288672

  11. PROSPECT: A Short-baseline Reactor Precision Spectrum and Oscillation Experiment

    NASA Astrophysics Data System (ADS)

    Langford, Thomas; Prospect Collaboration

    2015-10-01

    PROSPECT is a phased experiment consisting of segmented Li-loaded liquid scintillator antineutrino detectors designed to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. The experiment will be located at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. The first phase is a movable 2.5 tonne detector located 7-9 m from the compact, highly enriched uranium (HEU) core. Over the past two years, PROSPECT has deployed multiple prototype detectors at HFIR to understand the local background environment and demonstrate active and passive background rejection. Measuring the neutrino spectrum from 235U will give insight to the recent spectral discrepancies and provide an important benchmark for future reactor experiments. As a high statistics experiment, PROSPECT will probe the sterile neutrino best-fit region within one year of operation at HFIR.

  12. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    SciTech Connect

    Raussi, P.; Kainulainen, S.; Kouhia, J.

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  13. Analysis of the thorium axial blanket experiments in the PROTEUS reactor

    SciTech Connect

    White, J. R.; Ingersoll, D. T.; Schmocker, U.

    1980-01-01

    An extensive program of reactor physics experiments in GCFR fuel pin lattices has been completed recently at the PROTEUS critical facility located at EIR laboratory in Switzerland. The PROTEUS reactor consists of a central test zone surrounded by a uranium buffer and thermal driver region. The test lattices included a PuO/sub 2//UO/sub 2/ fuel region with internal and axial blankets of UO/sub 2/, ThO/sub 2/, and thorium metal. Detailed analysis of the thorium-bearing lattices has been performed at EIR and at ORNL in order to validate nuclear data and methods used for reactor physics analysis of advanced GCFR designs.

  14. Analysis of a boron-carbide-drum-controlled critical reactor experiment

    NASA Technical Reports Server (NTRS)

    Mayo, W. T.

    1972-01-01

    In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

  15. Neutron field for activation experiments in horizontal channel of training reactor VR-1

    NASA Astrophysics Data System (ADS)

    Stefanik, Milan; Katovsky, Karel; Vins, Miroslav; Soltes, Jaroslav; Zavorka, Lukas

    2014-11-01

    The experimental channels of nuclear reactors often serve for nuclear data measurement and validation. The dosimetry-foils activation technique was employed to measure neutron field parameters in the horizontal radial channel of the training reactor VR-1, and to test the possibility of using the reactor for scientific purposes. The reaction rates, energy spectral indexes, and neutron spectrum at several irradiation positions of the experimental channel were determined. The experimental results show the feasibility of the radial channel for irradiating experiments and open new possibilities for data validation by using this nuclear facility.

  16. Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

    NASA Technical Reports Server (NTRS)

    Fox, T. A.

    1973-01-01

    An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

  17. Geysering in boiling channels

    SciTech Connect

    Aritomi, Masanori; Takemoto, Takatoshi; Chiang, Jing-Hsien

    1995-09-01

    A concept of natural circulation BWRs such as the SBWR has been proposed and seems to be promising in that the primary cooling system can be simplified. The authors have been investigating thermo-hydraulic instabilities which may appear during the start-up in natural circulation BWRs. In our previous works, geysering was investigated in parallel boiling channels for both natural and forced circulations, and its driving mechanism and the effect of system pressure on geysering occurrence were made clear. In this paper, geysering is investigated in a vertical column and a U-shaped vertical column heated in the lower parts. It is clarified from the results that the occurrence mechanism of geysering and the dependence of system pressure on geysering occurrence coincide between parallel boiling channels in circulation systems and vertical columns in non-circulation systems.

  18. Pool boiling of nanofluids on rough and porous coated tubes: experimental and correlation

    NASA Astrophysics Data System (ADS)

    Cieśliński, Janusz T.; Kaczmarczyk, Tomasz Z.

    2014-06-01

    The paper deals with pool boiling of water-Al2O3 and water- Cu nanofluids on rough and porous coated horizontal tubes. Commercially available stainless steel tubes having 10 mm outside diameter and 0.6 mm wall thickness were used to fabricate the test heater. The tube surface was roughed with emery paper 360 or polished with abrasive compound. Aluminium porous coatings of 0.15 mm thick with porosity of about 40% were produced by plasma spraying. The experiments were conducted under different absolute operating pressures, i.e., 200, 100, and 10 kPa. Nanoparticles were tested at the concentration of 0.01, 0.1, and 1% by weight. Ultrasonic vibration was used in order to stabilize the dispersion of the nanoparticles. It was observed that independent of operating pressure and roughness of the stainless steel tubes addition of even small amount of nanoparticles augments heat transfer in comparison to boiling of distilled water. Contrary to rough tubes boiling heat transfer coefficient of tested nanofluids on porous coated tubes was lower compared to that for distilled water while boiling on porous coated tubes. A correlation equation for prediction of the average heat transfer coefficient during boiling of nanofluids on smooth, rough and porous coated tubes is proposed. The correlation includes all tested variables in dimensionless form and is valid for low heat flux, i.e., below 100 kW/m2.

  19. PR-EDB: Power Reactor Embrittlement Database - Version 3

    SciTech Connect

    Wang, Jy-An John; Subramani, Ranjit

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  20. Summary of the Workshop on Molten Salt Reactor Technologies Commemorating the 50th Anniversary of the Startup of the Molten Salt Reactor Experiment

    SciTech Connect

    Betzler, Benjamin R; Mays, Gary T

    2016-01-01

    A workshop on Molten Salt Reactor (MSR) technologies commemorating the 50th anniversary of the Molten Salt Reactor Experiment (MSRE) was held at Oak Ridge National Laboratory on October 15 16, 2015. The MSRE represented a pioneering experiment that demonstrated an advanced reactor technology: the molten salt eutectic-fueled reactor. A multinational group of more than 130 individuals representing a diverse set of stakeholders gathered to discuss the historical, current, and future technical challenges and paths to deployment of MSR technology. This paper provides a summary of the key messages from this workshop.

  1. Critical heat flux in subcooled flow boiling

    NASA Astrophysics Data System (ADS)

    Hall, David Douglas

    The critical heat flux (CHF) phenomenon was investigated for water flow in tubes with particular emphasis on the development of methods for predicting CHF in the subcooled flow boiling regime. The Purdue University Boiling and Two-Phase Flow Laboratory (PU-BTPFL) CHF database for water flow in a uniformly heated tube was compiled from the world literature dating back to 1949 and represents the largest CHF database ever assembled with 32,544 data points from over 100 sources. The superiority of this database was proven via a detailed examination of previous databases. The PU-BTPFL CHF database is an invaluable tool for the development of CHF correlations and mechanistic models that are superior to existing ones developed with smaller, less comprehensive CHF databases. In response to the many inaccurate and inordinately complex correlations, two nondimensional, subcooled CHF correlations were formulated, containing only five adjustable constants and whose unique functional forms were determined without using a statistical analysis but rather using the parametric trends observed in less than 10% of the subcooled CHF data. The correlation based on inlet conditions (diameter, heated length, mass velocity, pressure, inlet quality) was by far the most accurate of all known subcooled CHF correlations, having mean absolute and root-mean-square (RMS) errors of 10.3% and 14.3%, respectively. The outlet (local) conditions correlation was the most accurate correlation based on local CHF conditions (diameter, mass velocity, pressure, outlet quality) and may be used with a nonuniform axial heat flux. Both correlations proved more accurate than a recent CHF look-up table commonly employed in nuclear reactor thermal hydraulic computer codes. An interfacial lift-off, subcooled CHF model was developed from a consideration of the instability of the vapor-liquid interface and the fraction of heat required for liquid-vapor conversion as opposed to that for bulk liquid heating. Severe

  2. 76 FR 11524 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-02

    ... Boiling Water Reactors (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactors... participation in ACRS meetings were published in the Federal Register on October 21, 2010, (75 FR 65038-...

  3. 76 FR 18585 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-04

    ... Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor... October 21, 2010, (75 FR 65038- 65039). Detailed meeting agendas and meeting transcripts are available...

  4. Studying heat transfer enhancement for water boiling on a surface with micro- and nanorelief

    NASA Astrophysics Data System (ADS)

    Kuzma-Kichta, Yu. A.; Lavrikov, A. V.; Shustov, M. V.; Chursin, P. S.; Chistyakova, A. V.; Zvonarev, Yu. A.; Zhukov, V. M.; Vasil'eva, L. T.

    2014-03-01

    We present the results from a study of heat transfer enhancement for bulk water boiling at atmospheric pressure on a surface with micro- and nanorelief, including a relief formed from silicon carbide and aluminum oxide nanoparticles. Horizontally oriented steel tube 1.2 mm in diameter and copper plate 15 × 3 mm in size were selected as test sections. The process was recorded by means of a video camera, and the values of heat transfer, critical heat fluxes, and contact angles were measured. The use of surface with micro- and nanorelief makes it possible to obtain a significantly higher critical heat flux and boiling heat transfer coefficient owing to a change of surface wettability. The results of investigations can find use in compact heat exchangers, refrigerating plants, heat pipes, in the mirrors of high-capacity lasers, in the targets and resonators of charged particle accelerators and for external cooling of reactor vessels under emergency conditions.

  5. Boiling incipience and convective boiling of neon and nitrogen

    NASA Technical Reports Server (NTRS)

    Papell, S. S.; Hendricks, R. C.

    1977-01-01

    Forced convection and subcooled boiling heat transfer data for liquid nitrogen and liquid neon were obtained in support of a design study for a 30 tesla cryomagnet cooled by forced convection of liquid neon. This design precludes nucleate boiling in the flow channels as they are too small to handle vapor flow. Consequently, it was necessary to determine boiling incipience under the operating conditions of the magnet system. The cryogen data obtained over a range of system pressures, fluid flow rates, and applied heat fluxes were used to develop correlations for predicting boiling incipience and convective boiling heat transfer coefficients in uniformly heated flow channels. The accuracy of the correlating equations was then evaluated. A technique was also developed to calculate the position of boiling incipience in a uniformly heated flow channel. Comparisons made with the experimental data showed a prediction accuracy of plus or minus 15 percent

  6. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    SciTech Connect

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  7. Global analysis of bundle behavior in pressurized water reactor specific CORA experiments

    SciTech Connect

    Hering, W. ); Minato, Kazuo; Nagase, Fumihisa )

    1993-04-01

    At Kernforschungszentrum Karlsruhe, out-of-pile bundle experiments are performed in the CORA facility to investigate the behavior of light water reactor fuel elements during severe fuel damage accidents. To analyze the phenomena observed during the tests, such as claddin failure, oxidation, and deformation, as well as their influence on the post test bundle state, four pressurized water reactor specific tests are selected: CORA-2, CORA-3, CORA-5, and CORA-12. From each of these tests, a detailed global analysis using all the measured temperatures, pressures, and fluid compositions as well as videoscope information has been performed. To describe the post test bundle state quantitatively, axial profiles of the bundle cross-section area, the damage state of the rods, the average cladding oxidation, and the damage to the pellets are measured. The effects of CORA-specific components on the bundle melt progression and the measured axial profiles are identified and assessed. Most of the observations during the tests as well as the post test bundle state can be explained by the established common sequence of phenomena. For a better understanding of the melt progression, some physical phenomena, such as the energy release associated with the double-sided oxidation of the cladding, the melt release, or the melt relocation, must be analyzed in detail.

  8. Turning bubbles on and off during boiling using charged surfactants

    NASA Astrophysics Data System (ADS)

    Cho, H. Jeremy; Mizerak, Jordan P.; Wang, Evelyn N.

    2015-10-01

    Boiling--a process that has powered industries since the steam age--is governed by bubble formation. State-of-the-art boiling surfaces often increase bubble nucleation via roughness and/or wettability modification to increase performance. However, without active in situ control of bubbles, temperature or steam generation cannot be adjusted for a given heat input. Here we report the ability to turn bubbles `on and off' independent of heat input during boiling both temporally and spatially via molecular manipulation of the boiling surface. As a result, we can rapidly and reversibly alter heat transfer performance up to an order of magnitude. Our experiments show that this active control is achieved by electrostatically adsorbing and desorbing charged surfactants to alter the wettability of the surface, thereby affecting nucleation. This approach can improve performance and flexibility in existing boiling technologies as well as enable emerging or unprecedented energy applications.

  9. Boiling incipience and convective boiling of neon and nitrogen

    NASA Technical Reports Server (NTRS)

    Papell, S. S.; Hendricks, R. C.

    1977-01-01

    Forced convection and subcooled boiling heat transfer data for liquid nitrogen and liquid neon were obtained in support of a design study for a 30 tesla cryomagnet cooled by forced convection of liquid neon. The cryogen data obtained over a range of system pressures, fluid flow rates, and applied heat fluxes were used to develop correlations for predicting boiling incipience and convective boiling heat transfer coefficients in uniformly heated flow channels. The accuracy of the correlating equations was then evaluated. A technique was also developed to calculate the position of boiling incipience in a uniformly heated flow channel. Comparisons made with the experimental data showed a prediction accuracy of + or - 15 percent.

  10. Treatment of strong domestic sewage in a 96 m3 UASB reactor operated at ambient temperatures: two-stage versus single-stage reactor.

    PubMed

    Halalsheh, M; Sawajneh, Z; Zu'bi, M; Zeeman, G; Lier, J; Fayyad, M; Lettinga, G

    2005-03-01

    A 96 m3 UASB reactor was operated for 2.5 years under different conditions to assess the feasibility of treating strong sewage (COD(tot) = 1531 mg/l) at ambient temperatures with averages of 18 and 25 degrees C for winter and summer respectively. During the first year, the reactor was operated as a two-stage system at OLRs in the range of 3.6-5.0 kg COD/m3 d for the first stage and 2.9-4.6 kg COD/m3 d for the second stage. The results of the first stage showed average removals of 51% and 60% for COD(tot) and COD(ss) respectively without significant effect of temperature. The second stage reactor was unstable. The temperature affected sludge stabilization. During the second year, the first stage was operated as a single-stage UASB reactor at half of the previous loading rates. The results showed an average removal efficiency of 62% for COD(tot) during summer, while it dropped to 51% during wintertime. However, the effluent suspended solids were stabilized with VSS/TSS ratio around 0.50 all over the year. The sludge in the single-stage reactor was well-stabilized and exerted an excellent settlability. During the last three months of research, sludge was discharged regularly from the single-stage UASB reactor. The results showed no significant improvement in the performance in terms of COD(tot). Based on the results of the experiment, a single-stage UASB reactor operated at relatively long HRT is preferred above two-stage system at the Jordanian conditions. PMID:15501665

  11. Monte Carlo interpretation of the AMMON/REF experiment in EOLE for the JHR reactor calculations

    SciTech Connect

    Vaglio-Gaudard, C.; Leray, O.; Colombier, A. C.; D'Aletto, C.; Gaubert, L.; Gueton, O.; Hudelot, J. P.; Valentini, M.

    2012-07-01

    A new experiment, named AMMON and dedicated to the analysis of the JHR neutron physics and to the qualification of the associated neutron calculation tools, is currently in progress in the EOLE zero-power experimental reactor at CEA Cadarache. The first core configuration, so called 'reference configuration' (AMMON/REF) has already been loaded: it consists of an experimental zone of 7 JHR assemblies with U{sub 3}Si{sub 2} 27% {sup 235}U enriched fuel curved plates surrounded by a driver zone with 623 standard PWR UOx fuel pins. The analysis of the AMMON/REF measurements is presented in this paper. It is based on calculations performed with the 3-dimensional reference Monte Carlo TRIPOLI4.7 code and the JEFF3.1.1 European library: it highlights a good agreement between calculation and experiment concerning reactivity and power distribution in the experimental zone. Reactivity prediction is very satisfactory, despite the presence of a large aluminum quantity in the core: C-E=+ 376 pcm {+-} 290. The radial and axial plate power profiles are also in good agreement. Assembly power distribution is predicted within the experimental uncertainties (1{sigma}). (authors)

  12. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  13. Advanced Test Reactor In-Canal Ultrasonic Scanner: Experiment Design and Initial Results on Irradiated Plates

    SciTech Connect

    D. M. Wachs; J. M. Wight; D. T. Clark; J. M. Williams; S. C. Taylor; D. J. Utterbeck; G. L. Hawkes; G. S. Chang; R. G. Ambrosek; N. C. Craft

    2008-09-01

    An irradiation test device has been developed to support testing of prototypic scale plate type fuels in the Advanced Test Reactor. The experiment hardware and operating conditions were optimized to provide the irradiation conditions necessary to conduct performance and qualification tests on research reactor type fuels for the RERTR program. The device was designed to allow disassembly and reassembly in the ATR spent fuel canal so that interim inspections could be performed on the fuel plates. An ultrasonic scanner was developed to perform dimensional and transmission inspections during these interim investigations. Example results from the AFIP-2 experiment are presented.

  14. Characteristics of Pool Boiling on Graphite-Copper Composite Surfaces

    NASA Technical Reports Server (NTRS)

    Zhang, Nengli; Chao, David F.; Yang, Wen-Jei

    2002-01-01

    Nucleate pool boiling performance of different liquids on graphite-copper composite (Gr-Cu) surfaces has been experimentally studied and modeled. Both highly wetting fluids, such as freon-113 and pentane, and a moderately wetting fluid (water) were tested on the Gr-Cu surfaces with different graphite-fiber volume fractions to reveal the enhancement effects of the composite surfaces on the nucleate pool boiling. Results of the experiments show that the graphite-fiber volume fraction has an optimum value. The Gr-Cu composite surface with 25 percent graphite-fiber volume (f=0.25) has a maximum enhancement effect on the nucleate boiling heat transfer comparing to the pure copper surface. For the highly wetting fluid, the nucleate boiling heat transfer is generally enhanced on the Gr- Cu composite surfaces by 3 to 6 times shown. In the low heat flux region, the enhancement is over 6 times, but in the high heat flux region, the enhancement is reduced to about 40%. For the moderately wetting fluid (water), stronger enhancement of nucleate boiling heat transfer is achieved on the composite surface. It shown the experimental results in which one observes the nucleate boiling heat transfer enhancement of 5 to 10 times in the low heat flux region and an enhancement of 3 to 5 times in the high heat flux region. Photographs of bubble departure during the initial stage of nucleate boiling indicate that the bubbles detached from the composite surface are much smaller in diameter than those detached from the pure copper surface. Typical photographs are presented.It shows that the bubbles departed from the composite surface have diameters of only O(0.1) mm, while those departed from the pure copper surface have diameters of O(1) mm. It is also found that the bubbles depart from the composite surface at a much higher frequency, thus forming vapor columns. These two phenomena combined with high thermal conductivity of the graphite fiber are considered the mechanisms for such a

  15. Validation of KENO-VI: A comparison with hexagonal lattice light-water-reactor critical experiments

    SciTech Connect

    Lichtenwalter, J.J.

    1998-06-01

    The KENO-VI Monte Carlo code, released with Version 4.3 of the SCALE Code System, provides the capability to model more complex geometries than previously allowed by KENO-V.a. One significant improvement is the simplistic specification of hexprism unit cells and hexagonal arrays, an arduous task to complete in KENO-V.a. This report documents the validation of KENO-VI against 30 critical experiments consisting of low enriched uranium, light water reactor (LWR) fuel rods in hexagonal lattices with no poisons. The reference, enrichment, pitch, cladding, and core identification of the experiments are given. The results indicate that KENO-VI accurately calculates these critical experiments, with a bias of {minus}0.51% for the 238 group cross section library and {minus}0.24% for the 44 group cross section library. If these biases are properly taken into account, the KENO-VI code can be used with confidence for the design and safety analysis of storage and transportation systems of similar LWR type fuels.

  16. Analysis of two-phase flow phenomena with FLUENT-4 code in the experiments for advanced light water reactor safety

    SciTech Connect

    Miettinen, J.; Tuomainen, M.; Karppinen, I.; Tuunanen, J.

    2002-07-01

    In the development of advanced light water reactors, thermohydraulic phenomena are versatile in comparison with the present concepts. The new features are the passive safety systems, where energy transport takes place by natural circulation instead of forced flow. For cooling of the molten core, new concepts have been created including external vessel cooling and core catchers. In all new concepts, two-phase flow circulation patterns exist. The calculational tools should be capable of analysing multidimensional circulation created by the gravity field instead of the forced pump circulation. In spite of extensive model development for the one-dimensional Eulerian solutions for two-phase flow, multidimensional calculation is still a great challenge. The momentum transfer terms and turbulence models for the two-phase flow still require large efforts, although the turbulence models for the single phase flow are versatile and rather advanced at present. Two-phase models exist already now in several CFD codes. In VTT, most experience has been achieved with Fluent-4 Fluent-5 and at last Fluent-6 codes. Fluent-4 and Fluent-6 have the Euler-Euler solution for two-phase conservation equations, which is required for the flow conditions, where the volume fraction of both liquid and gas phases is important and the flow circulation is largely created by the gravity field. VTT is participating in several experimental projects on ALWRs, where multidimensional two-phase circulation is essential. This paper presents three examples of the use of CFD codes for analyses of ALWRs. The first example is connected with SWR 1000 reactor form Framatome ANP. Framatome ANP is performing experiments for evaluation of external cooling of the Reactor Pressure Vessel (RPV) of SWR 1000. The experiments are aimed for determining the limits to avoid critical heat fluxes (CHFs). The experimental programme is carried out in three steps. The first part, the air-water experiments, has been analysed at

  17. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    SciTech Connect

    Tomberlin, Terry Alan

    2002-06-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to "major modifications" and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

  18. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    SciTech Connect

    Tomberlin, T.A.

    2002-06-19

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

  19. Safety in the ARIES-3 D-He-3 tokamak reactor design

    NASA Astrophysics Data System (ADS)

    Herring, J. S.; Dolan, T. J.

    1991-11-01

    The ARIES-3 reactor study is an extensive examination of the viability of a D-He-3 He-fueled commercial tokamak power reactor. Because neutrons are produced only through side reactions, the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The ARIES-3 reactor uses an organic coolant to cool the first wall, shield and divertor. The organic coolant has a low vapor pressure at the operating temperature required for good thermal efficiency. Radiation damage requires processing the coolant to remove and crack radiolytic products that would otherwise foul cooling surfaces. The cracking process produces waste, which must be disposed of through incineration or burial. We estimated the offsite doses due to incineration at five candidate locations. The plasma confinement requirements for a D-He-3 reactor are much more challenging than those for a D-T reactor. Thus, the demands on the divertor are more severe, particularly during a disruption. We explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. We also modeled a loss-of-cooling accident in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. We analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps. For our reference design, the tritium will be burned in the plasma. These results re-emphasize the need for low activation materials and advanced divertor designs, even in reactors using advanced fuels.

  20. 3D Computations and Experiments

    SciTech Connect

    Couch, R; Faux, D; Goto, D; Nikkel, D

    2004-04-05

    This project consists of two activities. Task A, Simulations and Measurements, combines all the material model development and associated numerical work with the materials-oriented experimental activities. The goal of this effort is to provide an improved understanding of dynamic material properties and to provide accurate numerical representations of those properties for use in analysis codes. Task B, ALE3D Development, involves general development activities in the ALE3D code with the focus of improving simulation capabilities for problems of mutual interest to DoD and DOE. Emphasis is on problems involving multi-phase flow, blast loading of structures and system safety/vulnerability studies.

  1. 3D Computations and Experiments

    SciTech Connect

    Couch, R; Faux, D; Goto, D; Nikkel, D

    2003-05-12

    This project is in its first full year after the combining of two previously funded projects: ''3D Code Development'' and ''Dynamic Material Properties''. The motivation behind this move was to emphasize and strengthen the ties between the experimental work and the computational model development in the materials area. The next year's activities will indicate the merging of the two efforts. The current activity is structured in two tasks. Task A, ''Simulations and Measurements'', combines all the material model development and associated numerical work with the materials-oriented experimental activities. Task B, ''ALE3D Development'', is a continuation of the non-materials related activities from the previous project.

  2. Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

    SciTech Connect

    Higgins, J.; Nasta, K.

    1997-01-01

    The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs.

  3. How Does Water Boil?

    NASA Astrophysics Data System (ADS)

    Zahn, Dirk

    2004-11-01

    Insight into the boiling of water is obtained from molecular dynamics simulations. The process is initiated by the spontaneous formation of small vacuum cavities in liquid water. By themselves, these defects are very short lived. If, however, several cavities occur at close distances, they are likely to merge into larger vacuum holes. At the liquid-vapor interfaces, single or small groups of water molecules tend to leave the liquid surface. Once the system is propagated beyond the transition state, these evaporation events outnumber the competing reintegration into the hydrogen-bonded network.

  4. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    SciTech Connect

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant.

  5. NEBA-3 pin and cermet hybrid bimodal reactor

    SciTech Connect

    Weitzberg, A.; Josloff, A.T.; Mondt, J.F.

    1995-12-31

    Early in 1994 a bimodal (power and propulsion) reactor concept that made use of two different fuel types was identified. UN pin fuel was selected for long lifetime to produce electricity, and UO{sub 2}-W cermet was selected for short-time high-temperature propulsion. During the last year the concept has been refined and analyzed as part of the DOE contribution to the joint Air Force Phillips Laboratory-Department of Energy Bimodal Program. With the exception of refractory metal vacuum gaps between the components containing hydrogen propellant and the refractory metal components containing lithium, and high temperature liquid metal-gas heat exchangers, the concept identified as Nuclear Engine for Bimodal Applications (NEBA)-3 uses previously developed and demonstrated technologies, including high efficiency closed Brayton cycle turboalternator-compressors for power production. Using near-term technology NEBA-3 can, as an upper stage to an Atlas IIAS launch vehicle, provide propulsion with specific impulse of about 850 seconds at thrusts of 90 to 925 Newtons, and 10 kilowatts of electricity for ten years. Based on previous cost estimates by DOE and NASA, a minimum program could deliver a flight system such as NEBA-3 within six years at a cost of from $500 to $700 million. The US space reactor power and propulsion community now has the technical capability to deliver such a system early in the next century.

  6. Two-dimensional modeling of sodium boiling in a simulated LMFBR loss-of-flow test

    SciTech Connect

    Rose, S.D.

    1984-01-01

    Loss-of-flow (LOF) accidents are of major importance in LMFBR safety. Tests have been performed to simulate the simultaneous failure of all primary pumps and reactor shutdown systems in a 37-pin electrically heated test bundle installed in the KNS sodium boiling loop at the Institute of Reactor Development, Karlsruhe. The tests simulated LOF conditions of the German prototype LMFBR, the SNR 300. The main objectives of these tests were to characterize the transient boiling development to cladding dryout and to provide data for validation of sodium boiling codes. One particular LOF test, designated L22, at full power was selected as a benchmark exercise for comparison of several codes at the Eleventh Meeting of the Liquid Metal Boiling Working Group (LMBWG) held in Grenoble, France, in October 1984. In this paper, the results of the calculations performed at ORNL with the two-dimensional (2-D) boiling code THORAX are presented.

  7. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  8. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  9. (Investigation of subcooled hydrothermal boiling in ground water flow channels as a source of harmonic tremors)

    SciTech Connect

    Not Available

    1989-01-01

    As a first step toward assessing the ability of hydrothermal boiling to explain geothermal ground noise and volcanic tremor observations, we are investigating the acoustic power spectrum of boiling (the source'' spectrum in the above model). We simulate boiling in the lab by injecting high pressure steam from a boiler into a pressure vessel filled with water. The water pressure fluctuations that result from the repeated formation and collapse of steam bubbles at the steam inlet vents are recorded by a hydrophone whose output is digitized at 2 {times} 10{sup 4} samples/second by a computer. The range of pressure and temperature conditions attainable within the pressure vessel is limited to <3.5 bars, <139{degree}C, due to the finite strength of observation windows affixed to the pressure vessel. Therefore, dimensional analysis will be used to correlate the experimental results with the pertinent experimental variables. Besides the overall shape of the boiling power spectrum, we are investigating the absolute spectral levels in frequency bands typical of geothermal ground noise and volcanic tremor (0.5 Hz-10 Hz), and the ratio of acoustic power liberated to total available power. The values of these parameters are critical to hydrothermal boiling's ability to generate ground motion amplitudes in accordance with observation. If it can be shown that the range of observed ground noise/tremor amplitudes can be accounted for by hydrothermal boiling at reasonable heat transfer rates, this knowledge would be invaluable to designers of seismic monitoring experiments who are interested in geothermal resource exploration/evaluation and volcanic eruption prediction.

  10. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  11. Calculated k-effectives for light water reactor typical, U + Pu nitrate solution critical experiments

    SciTech Connect

    Primm, R.T. III; Mincey, J.F.

    1982-01-01

    The Department of Energy's Consolidated Fuel Reprocessing Program has as a goal the design of nuclear fuel reprocessing equipment. In order to validate computer codes used for criticality analyses in the design of such equipment, k-effectives have been calculated for several U + Pu nitrate solution critical experiments. As of January 1981, descriptions of 45 unpoisoned, U + Pu solution experiments were available in the open literature. Twelve of these experiments were performed with solutions which have physical characteristics typical of dissolved, light water reactor fuel. This paper contains a discussion of these twelve experiments, a review of the calculational procedure used to determine k-effectives, and the results of the calculations.

  12. Boiling Fluids Behave Quite Differently in Space

    NASA Video Gallery

    The boiling process is really different in space, since the vapor phase of a boiling liquid does not rise via buoyancy. Spacecraft and Earth-based systems use boiling to efficiently remove large am...

  13. Tank Pressure Control Experiment: Thermal Phenomena in Microgravity. Video 3 of 4

    NASA Technical Reports Server (NTRS)

    1996-01-01

    The report presents the results of the flight experiment Tank Pressure Control Experiment/Thermal Phenomena (TPCE/TP) performed in the microgravity environment of the space shuttle. TPCE/TP, flown on the Space Transportation System STS-52, was a second flight of the Tank Pressure Control Experiment (TPCE). The experiment used Freon 113 at near saturation conditions. The test tank was filled with liquid to about 83 percent by volume. The experiment consisted of 21 tests. Each test generally started with a heating phase to increase the tank pressure and to develop temperature stratification in the fluid, followed by a fluid mixing phase for the tank pressure reduction and fluid temperature equilibration. The heating phase provided pool boiling data from large (relative to bubble sizes) heating surfaces (0.1046 m by 0.0742 m) at low heat fluxes (0.23 to 1.16 kW/m(exp 2)). The system pressure and the bulk liquid subcooling varied from 39 to 78 kPa and 1 to 3 deg C, respectively. The boiling process during the entire heating period, as well a jet-induced mixing process for the first 2 min. of the mixing period, was also recorded on video. Analyses of data from the two flight experiments (TPCE and TPCE/TP) and their comparison with the results obtained in drop tower experiments suggest that as Bond number approaches zero the flow pattern produced by an axial jet and the mixing time can be predicted by the Weber number. This is video 3 of 4.

  14. A Fundamental Study of Nucleate Pool Boiling Under Microgravity

    NASA Technical Reports Server (NTRS)

    Ervin, Jamie S.; Merte, Herman, Jr.

    1996-01-01

    An experimental study of incipient boiling in short-term microgravity and with a/g = +/- 1 for pool boiling was performed. Calibrated thin gold films sputtered on a smoothly polished quartz surface were used simultaneously for thermal-resistance measurements and heating of the boiling surface. The gold films were used for both transient and quasi-steady heating surface temperature measurements. Two test vessels were constructed for precise measurement and control of fluid temperature and pressure: a laboratory pool boiling vessel for the a/g = +/- 1 experiments and a pool boiling vessel designed for the 131 m free-fall in the NASA Lewis Research Center Microgravity Research Facility for the microgravity tests. Measurements included the heater surface temperature, the pressure near the heating surface, the bulk liquid temperatures. High speed photography (up to 1,000 frames per second) was used in the experiments. With high quality microgravity and the measured initial temperature of the quiescent test fluid, R113, the temperature distribution in the liquid at the moment of boiling inception resulting from an imposed step in heat flux is known with a certainty not possible previously. The types of boiling propagation across the large flat heating surface, some observed here for the first time, are categorized; the conditions necessary for their occurrence are described. Explosive boiling propagation with a striking pattern of small scale protuberances over the entire vapor mass periphery not observed previously at low heat flux levels (on the order of 5 W/cm(exp 2)) is described. For the heater surface with a/g = -1, a step in the heater surface temperature of short duration was imposed. The resulting liquid temperature distribution at the moment of boiling inception was different from that obtained with a step in heat flux.

  15. A fundamental study of nucleate pool boiling under microgravity

    NASA Technical Reports Server (NTRS)

    Ervin, Jamie S.; Merte, Herman, Jr.

    1991-01-01

    An experimental study of incipient boiling in short-term microgravity and with a/g = +/- 1 for pool boiling was performed. Calibrated thin gold films sputtered on a smoothly polished quartz surface were used simultaneously for thermal resistance measurements and heating of the boiling surface. The gold films were used for both transient and quasi-steady heating surface temperature measurements. Two test vessels were constructed for precise measurement and control of fluid temperature and pressure: a laboratory pool boiling vessel for the a/g = +/- experiments and a pool boiling vessel designed for the 131 m free-fall in the NASA Lewis Research Center Microgravity Research Facility for the microgravity tests. Measurements included the heater surface temperature, the pressure near the heating surface, and the bulk liquid temperatures. High speed photography was used in the experiments. With high quality microgravity and the measured initial temperature of the quiescent test fluid, R113, the temperature distribution in the liquid at the moment of boiling inception resulting from an imposed step in heat flux is known with a certainty not possible previously. The types of boiling propagation across the large flat heating surface are categorized; the conditions necessary for their occurrence are described. Explosive boiling propagation with a striking pattern of small scale protuberances over the entire vapor mass periphery not observed previously at low heat flux levels is described. For the heater surface with a/g = -1, a step in the heater surface temperature of short duration was imposed. The resulting liquid temperature distribution at the moment of boiling inception was different from that obtained with a step in heat flux.

  16. Circulation system for flowing uranium hexafluoride cavity reactor experiments

    NASA Technical Reports Server (NTRS)

    Jaminet, J. F.; Kendall, J. S.

    1976-01-01

    Research related to determining the feasibility of producing continuous power from fissile fuel in the gaseous state is presented. The development of three laboratory-scale flow systems for handling gaseous UF6 at temperatures up to 500 K, pressure up to approximately 40 atm, and continuous flow rates up to approximately 50g/s is presented. A UF6 handling system fabricated for static critical tests currently being conducted is described. The system was designed to supply UF6 to a double-walled aluminum core canister assembly at temperatures between 300 K and 400 K and pressure up to 4 atm. A second UF6 handling system designed to provide a circulating flow of up to 50g/s of gaseous UF6 in a closed-loop through a double-walled aluminum core canister with controlled temperature and pressure is described. Data from flow tests using UF6 and UF6/He mixtures with this system at flow rates up to approximately 12g/s and pressure up to 4 atm are presented. A third UF6 handling system fabricated to provide a continuous flow of UF6 at flow rates up to 5g/s and at pressures up to 40 atm for use in rf-heated, uranium plasma confinement experiments is described.

  17. Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

  18. The first critical experiment with a LEU Russian fuel IRT-4M at the training reactor VR-1

    SciTech Connect

    Frybort, Jan

    2008-07-15

    A critical experiment is a standard part of training of students at the Training Reactor VR-1 operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague. In autumn 2005 the HEU fuel IRT-3M with enrichment 36 % {sup 235}U was replaced by the LEU fuel IRT-4M with enrichment 19.7 % {sup 235}U. The fuel replacement at the VR-1 Reactor is a part of an international program RERTR. This Paper presents basic information about preparation for the fuel replacement and approaching of the first critical state with the new zone configuration C1 which replaced B1 core with the old IRT-3M fuel. The whole process was carried out according to the Czech law and the relevant international recommendations. The experience with the VR-1 operation confirms the assumption that the C1 core configuration will be suitable from the point of view of the reactivity balance for the long term safe operation of the Training Reactor VR-1. (author)

  19. Materials experience and selection for nuclear materials production reactor heat exchangers

    SciTech Connect

    Marra, J.E.; Louthan, M.R. Jr.

    1990-01-01

    The primary coolant systems for the heavy-water nuclear materials production reactors at the Savannah River Site are coupled to the secondary coolant systems through shell and tube heat exchangers. The head, shell, and tube sheets of these heat exchangers are fabricated from AISI Type 304 grades of austenitic stainless steel. The 8,957 tubes in each heat exchanger were originally fabricated from Type 304 stainless steel, but service experience has lead to the use of Sea Cure tubing in newer systems. The design includes double tube sheets, core rods, and 33,410 square feet of heat transfer surface. Tubes are rolled into the tube sheets and seal welded after rolling. The tubes contain Type 304 stainless steel rods which are positioned in the center of each tube axis to increase the fraction of the cooling water contacting the heat transfer surface. Each reactor utilizes twelve heat exchangers; thus the 120+ reactor-years of operating experience provide approximately 1,440 heat exchanger-years of service. Fatigue, stress corrosion cracking, crevice corrosion, and pitting have been observed during the service life. This paper describes the observed degradation processes and uses the operational experience to recommend materials for the Heavy Water -- New Production Reactor (HW-NPR).

  20. Transition boiling heat transfer and the film transition regime

    NASA Technical Reports Server (NTRS)

    Ramilison, J. M.; Lienhard, J. H.

    1987-01-01

    The Berenson (1960) flat-plate transition-boiling experiment has been recreated with a reduced thermal resistance in the heater, and an improved access to those portions of the transition boiling regime that have a steep negative slope. Tests have been made in Freon-113, acetone, benzene, and n-pentane boiling on horizontal flat copper heaters that have been mirror-polished, 'roughened', or teflon-coated. The resulting data reproduce and clarify certain features observed by Berenson: the modest surface finish dependence of boiling burnout, and the influence of surface chemistry on both the minimum heat flux and the mode of transition boiling, for example. A rational scheme of correlation yields a prediction of the heat flux in what Witte and Lienhard (1982) previously identified as the 'film-transition boiling' region. It is also shown how to calculate the heat flux at the boundary between the pure-film, and the film-transition, boiling regimes, as a function of the advancing contact angle.

  1. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  2. Sequential UASB and dual media packed-bed reactors for domestic wastewater treatment - experiment and simulation.

    PubMed

    Rodríguez-Gómez, Raúl; Renman, Gunno

    2016-01-01

    A wastewater treatment system composed of an upflow anaerobic sludge blanket (UASB) reactor followed by a packed-bed reactor (PBR) filled with Sorbulite(®) and Polonite(®) filter material was tested in a laboratory bench-scale experiment. The system was operated for 50 weeks and achieved very efficient total phosphorus (P) removal (99%), 7-day biochemical oxygen demand removal (99%) and pathogenic bacteria reduction (99%). However, total nitrogen was only moderately reduced in the system (40%). A model focusing on simulation of organic material, solids and size of granules was then implemented and validated for the UASB reactor. Good agreement between the simulated and measured results demonstrated the capacity of the model to predict the behaviour of solids and chemical oxygen demand, which is critical for successful P removal and recovery in the PBR. PMID:27332842

  3. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    SciTech Connect

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2012-07-01

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

  4. Theoretical and experimental study of inverted annular film boiling and regime transition during reflood transients

    NASA Astrophysics Data System (ADS)

    Mohanta, Lokanath

    The Loss of Coolant Accident (LOCA) is a design basis accident for light water reactors that usually determines the limits on core power. During a LOCA, film boiling is the dominant mode of heat transfer prior to the quenching of the fuel rods. The study of film boiling is important because this mode of heat transfer determines if the core can be safely cooled. One important film boiling regime is the so-called Inverted Annular Film Boiling (IAFB) regime which is characterized by a liquid core downstream of the quench front enveloped by a vapor film separating it from the fuel rod. Much research have been conducted for IAFB, but these studies have been limited to steady state experiments in single tubes. In the present work, subcooled and saturated IAFB are investigated using high temperature reflood data from the experiments carried out in the Rod Bundle Heat Transfer (RBHT) test facility. Parametric effects of system parameters including the pressure, inlet subcooling, and flooding rate on the heat transfer are investigated. The heat transfer behavior during transition to Inverted Slug Film Boiling (ISFB) regime is studied and is found to be different than that reported in previous studies. The effects of spacer grids on heat transfer in the IAFB and ISFB regimes are also presented. Currently design basis accidents are evaluated with codes in which heat transfer and wall drag must be calculated with local flow parameters. The existing models for heat transfer are applicable up to a void fraction of 0.6, i.e. in the IAFB regime and there is no heat transfer correlation for ISFB. A new semi-empirical heat transfer model is developed covering the IAFB and ISFB regimes which is valid for a void fraction up to 90% using the local flow variables. The mean absolute percentage error in predicting the RBHT data is 11% and root mean square error is 15%. This new semi-empirical model is found to compare well with the reflood data of FLECHT-SEASET experiments as well as data

  5. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    SciTech Connect

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  6. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  7. Loss-of-coolant accident experiment at the AVR gas-cooled reactor

    SciTech Connect

    Cleveland, J.; Krueger, K.; Kernforschungsanlage Juelich G.m.b.H. . Arbeitsgemeinschaft Versuchsreaktor)

    1989-01-01

    Loss-of-coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small High-Temperature Gas Cooled Reactor (HTGR) designs, loss-of-coolant accident (LOCA) simulation tests have been conducted with the German pebble-bed High-Temperature Reactor AVR. The AVR is the only nuclear power plant ever to have been intentionally subjected to LOCA conditions. The LOCA test was planned to create conditions that would exist if a rapid LOCA occurred with the reactor operating at full power. The tests demonstrated this reactor's safe response to an accident in which the coolant escapes from the reactor core and no emergency system is available to provide coolant flow to the core. The test is of special interest because it demonstrates the inherent safety features incorporated into modular HTGR designs. The main LOCA test lasted for 5 d. After the test began, core temperatures increased for {approximately}13 h and then gradually and continually decreased as the rate of heat dissipation from the core exceeded accident levels of decay power. Throughout the test, temperatures remained below limiting values for the core and other reactor components. 3 refs., 9 figs., 1 tab.

  8. Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

    SciTech Connect

    Boardman, J.R.

    1994-10-01

    This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures.

  9. Impact of transmutation issues on interpretation of data obtained from fast reactor irradiation experiments

    NASA Astrophysics Data System (ADS)

    Greenwood, L. R.; Garner, F. A.

    2004-08-01

    The subject of fission-fusion correlation is usually cast in terms of reactor-to-reactor differences, but recently the fusion community has become aware of the impact of differences within a given surrogate facility, especially in constant time experiments when different dose levels are attained in different positions of one reactor. For some materials, it is not safe to assume that in-reactor spectral variations are small and of no consequence. This point is illustrated using calculations for fusion-relevant materials that were irradiated in the Fast Flux Test Facility-Materials Open Test Assembly (FFTF-MOTA) over a wide range of in-core and out-of-core positions spanning more than two orders of magnitude in dpa rate. It is shown that although both the neutron spectrum and flux changes, the spectral effectiveness factor, dpa/10 22 n/cm 2 ( E>0.1 MeV), remains remarkably constant over this range. The transmutation rate per dpa varies strongly with reactor position, however.

  10. Impact of Transmutation Issues on Interpretation of Data Obtained From Fast Reactor Irradiation Experiments

    SciTech Connect

    Greenwood, Lawrence R.; Garner, Francis A.

    2004-04-15

    The subject of fission-fusion correlation is usually cast in terms of reactor-to-reactor differences, but recently the fusion community has become aware of the impact of differences within a given surrogate facility, especially in constant time experiments when different dose levels are attained in different positions of one reactor. For some materials, it is not safe to assume that in-reactor spectral variations are small and of no consequence. This point is illustrated using calculations for fusion-relevant materials that were irradiated in the Fast Flux Test Facility – Materials Open Test Assembly (FFTF-MOTA) over a wide range of in-core and out of core positions spanning more than two orders of magnitude in dpa rate. It is shown that although both the neutron spectrum and flux changes, the spectral effectiveness factor, dpa/10(22) superscript n/cm(2) superscript (E > 0.1 MeV), remains remarkably constant over this range. The transmutation rate per dpa varies strongly with reactor position, however.

  11. Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

    NASA Astrophysics Data System (ADS)

    Žagar, Tomaž; Božič, Matjaž; Ravnik, Matjaž

    2004-12-01

    In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived (γ emitting) radioactive nuclides in the concrete were found to be 133Ba, 60Co and 152Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jožef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. 133Ba, 41Ca) are not included in the IAEA and EU basic safety standards.

  12. Film boiling heat transfer from a sphere in natural and forced convection of freon-113

    SciTech Connect

    Dix, D.; Orozco, J. )

    1990-01-01

    Boiling heat transfer fluxes were measured on a 3.84-cm hollow copper sphere, in both forced convection and pool boiling, as a function of angular position in Freon 113. This paper reports on forced-convection tests run at speeds of 0.5 to 1.9 m/s. These tests were conducted in the stable film boiling region of the boiling curve. Significant heat transfer rates were measured in the vapor wake region of the sphere for flow film boiling. Video observations of the boiling process revealed that the flow film boiling vapor removal mechanism consisted of periodic formation and detachment of a vapor wake in the rear of the sphere. For pool boiling it was found that the heated surface had a uniform rate of energy dissipation in the stable film boiling regime, whereas in forced convection the film boiling rate was dependent on angular position. Pool film boiling tests also showed multiple humps (more than one maximum heat flux) in the boiling curve when the liquid was subcooled.

  13. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    SciTech Connect

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  14. Theoretical analysis of the subcritical experiments performed in the IPEN/MB-01 research reactor facility

    SciTech Connect

    Lee, S. M.; Dos Santos, A.

    2012-07-01

    The theoretical analysis of the subcritical experiments performed at the IPEN/MB-01 reactor employing the coupled NJOY/AMPX-II/TORT systems was successfully accomplished. All the analysis was performed employing ENDF/B-VII.0. The theoretical approach follows all the steps of the subcritical model of Gandini and Salvatores. The theory/experiment comparison reveals that the calculated subcritical reactivity is in a very good agreement to the experimental values. The subcritical index ({xi}) shows some discrepancies although in this particular case some work still have to be made to model in a better way the neutron source present in the experiments. (authors)

  15. Spacelab 3 vapor crystal growth experiment

    NASA Technical Reports Server (NTRS)

    Schnepple, W.; Vandenberg, L.; Skinner, N.; Ortale, C.

    1987-01-01

    The Space Shuttle Challenger, with Spacelab 3 as its payload, was launched into orbit April 29, 1985. The mission, number 51-B, emphasized materials processing in space, although a wide variety of experiments in other disciplines were also carried onboard. One of the materials processing experiments on this flight is described, specifically the growth of single crystals of mercuric iodide by physical vapor transport.

  16. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  17. Assessment of Biasi and Columbia University CHF correlations with GE 3x3 rod bundle experiment. [PWR; BWR

    SciTech Connect

    Chen, B.C.J.; Chien, T.H.; Sha, W.T.; Kim, J.H.

    1984-01-01

    The critical heat flux (CHF), at which a sudden degradation of heat transfer occurs without corresponding decrease in heat generation, is one of the limiting parameters for safe operation of nuclear reactors. Reactor operation beyond the CHF causes a rapid rise in fuel cladding temperature and thus should be avoided to maintain the fuel element integrity. Reactor power limits are therefore set so that a prescribed safety margin below the CHF is maintained. Two CHF correlations are evaluated for reactor core thermal hydraulic analysis: the Biasi correlation and the Columbia University correlation. The BODYFIT-2PE computer code is used for this assessment. The CHF predicted by the BODYFIT-2PE using the two correlations is compared with GE 3x3 rod bundle CHF experiment.

  18. The Influence of the Heating Condition on the Void Fraction in a Boiling Channel

    NASA Astrophysics Data System (ADS)

    Umekawa, H.; Nakamura, S.; Fujiyoshi, S.; Ami, T.; Ozawa, M.; Saito, Y.; Ito, D.

    The void fraction profile in a boiling channel is essential in analyzing convective flow boiling, where several investigations have been conducted. But due to the difficulty in the treatment of the non-thermodynamic equilibrium phenomena under subcooled conditions, the issues in comprehensive void fraction profile has not been solved, yet. To improve the understanding of these phenomena, detailed measurement results are required. In this investigation, by using five kinds of test sections, i.e. I.D.=3 mm L=400 mm, I.D.=5 mm L=200, 400, 1000 mm, and I.D.=10 mm L=400 mm, the void fraction was measured quantitatively over the whole length. For the measurements, thermal neutron radiography at the B-4 port of the Kyoto University Research Reactor was used. This facility is designed for the visualization of forced convective flow boiling in vertical tubes. To introduce the performance of this facility, this paper presents the measurement results of void fraction and the estimation results briefly.

  19. Film boiling of mercury droplets

    NASA Technical Reports Server (NTRS)

    Baumeister, K. J.; Schoessow, G. J.; Chmielewski, C. E.

    1975-01-01

    Vaporization times of mercury droplets in Leidenfrost film boiling on a flat horizontal plate are measured in an air atmosphere. Extreme care was used to prevent large amplitude droplet vibrations and surface wetting; therefore, these data can be compared to film boiling theory. For these data, diffusion from the upper surface of the drop is a dominant mode of mass transfer from the drop. A closed-form analytical film boiling theory is developed to account for the diffusive evaporation. Reasonable agreement between data and theory is seen.

  20. Film boiling of mercury droplets

    NASA Technical Reports Server (NTRS)

    Baumeister, K. J.; Schoessow, G. J.; Chmielewski, C. E.

    1975-01-01

    Vaporization times of mercury droplets in Leidenfrost film boiling on a flat horizontal plate are measured in an air atmosphere. Extreme care was used to prevent large amplitude droplet vibrations and surface wetting; therefore, these data can be compared to film boiling theory. Diffusion from the upper surface of the drop appears as a dominant mode of mass transfer from the drop. A closed-form analytical film boiling theory is developed to account for the diffusive evaporation. Reasonable agreement between data and theory is seen.

  1. Elucidating reactivity regimes in cyclopentane oxidation: Jet stirred reactor experiments, computational chemistry, and kinetic modeling

    DOE PAGESBeta

    Al Rashidi, Mariam J.; Thion, Sebastien; Togbe, Casimir; Dayma, Guillaume; Mehl, Marco; Dagaut, Philippe; Pitz, William J.; Zador, Judit; Sarathy, S. Mani

    2016-05-01

    This study is concerned with the identification and quantification of species generated during the combustion of cyclopentane in a jet stirred reactor (JSR). Experiments were carried out for temperatures between 740 and 1250 K, equivalence ratios from 0.5 to 3.0, and at an operating pressure of 10 atm. The fuel concentration was kept at 0.1% and the residence time of the fuel/O2/N2 mixture was maintained at 0.7 s. The reactant, product, and intermediate species concentration profiles were measured using gas chromatography and Fourier transform infrared spectroscopy. The concentration profiles of cyclopentane indicate inhibition of reactivity between 850-1000 K for φ=2.0more » and φ=3.0. This behavior is interesting, as it has not been observed previously for other fuel molecules, cyclic or non-cyclic. A kinetic model including both low- and high-temperature reaction pathways was developed and used to simulate the JSR experiments. The pressure-dependent rate coefficients of all relevant reactions lying on the PES of cyclopentyl + O2, as well as the C-C and C-H scission reactions of the cyclopentyl radical were calculated at the UCCSD(T)-F12b/cc-pVTZ-F12//M06-2X/6-311++G(d,p) level of theory. The simulations reproduced the unique reactivity trend of cyclopentane and the measured concentration profiles of intermediate and product species. Furthermore, sensitivity and reaction path analyses indicate that this reactivity trend may be attributed to differences in the reactivity of allyl radical at different conditions, and it is highly sensitive to the C-C/C-H scission branching ratio of the cyclopentyl radical decomposition.« less

  2. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

  3. Boiling local heat transfer enhancement in minichannels using nanofluids

    PubMed Central

    2013-01-01

    This paper reports an experimental study on nanofluid convective boiling heat transfer in parallel rectangular minichannels of 800 μm hydraulic diameter. Experiments are conducted with pure water and silver nanoparticles suspended in water base fluid. Two small volume fractions of silver nanoparticles suspended in water are tested: 0.000237% and 0.000475%. The experimental results show that the local heat transfer coefficient, local heat flux, and local wall temperature are affected by silver nanoparticle concentration in water base fluid. In addition, different correlations established for boiling flow heat transfer in minichannels or macrochannels are evaluated. It is found that the correlation of Kandlikar and Balasubramanian is the closest to the water boiling heat transfer results. The boiling local heat transfer enhancement by adding silver nanoparticles in base fluid is not uniform along the channel flow. Better performances and highest effect of nanoparticle concentration on the heat transfer are obtained at the minichannels entrance. PMID:23506445

  4. Turning bubbles on and off during boiling using charged surfactants

    PubMed Central

    Cho, H. Jeremy; Mizerak, Jordan P.; Wang, Evelyn N.

    2015-01-01

    Boiling—a process that has powered industries since the steam age—is governed by bubble formation. State-of-the-art boiling surfaces often increase bubble nucleation via roughness and/or wettability modification to increase performance. However, without active in situ control of bubbles, temperature or steam generation cannot be adjusted for a given heat input. Here we report the ability to turn bubbles ‘on and off' independent of heat input during boiling both temporally and spatially via molecular manipulation of the boiling surface. As a result, we can rapidly and reversibly alter heat transfer performance up to an order of magnitude. Our experiments show that this active control is achieved by electrostatically adsorbing and desorbing charged surfactants to alter the wettability of the surface, thereby affecting nucleation. This approach can improve performance and flexibility in existing boiling technologies as well as enable emerging or unprecedented energy applications. PMID:26486275

  5. NRC review of the CANDU-3 reactor design

    SciTech Connect

    Kennedy, J.L. )

    1993-01-01

    This paper presents an overview of the US Nuclear Regulatory Commission's (NRC's) effort to complete an early review of the Canada deuterium uranium (CANDU)-3 reactor design prior to formal submittal of an application for standard design certification. The NRC is conducting a review of the CANDU-3 design in support of a request by AECL Technologics, the US sponsor of the design. The purpose of this review is to encourage early interactions by applicants, vendors, and government agencies with the NRC. The CANDU-3 design is being developed by Atomic Energy of Canada, Limited (AECL), whose CANDU operations are based in Mississauga, Ontario. AECL Technologies, a US subsidiary of AECL, Incorporated, informed the NRC of its intent to seek design certification of the CANDU-3 design under the provisions of 10CFR52 in a letter to the NRC dated May 25, 1989. This paper describes the commission's basis for this type of early review, its purposes and objectives, key elements of the review, the intended product, and the schedule.

  6. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    SciTech Connect

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L.; Matsui, H.

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  7. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  8. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  9. Analysis of the DHCE experiment in the position A10 of the ATR reactor

    SciTech Connect

    Gomes, I.C.; Smith, D.L.; Tsai, H.

    1997-08-01

    Calculations were performed to assess the possibility of performing DHCE experiments in mixed spectrum fission reactors. Calculated values of key parameters were compared with limit values for each quantity. The values calculated were: He-4 production from the {sup 6}Li(n,t){sup 4}He reaction, tritium leakage, required tritium concentration in lithium, initial tritium charge per capsule, and helium to dpa ratio after 10 dpa of irradiation.

  10. Study of in-reactor creep of vanadium alloy in the HFIR RB-12J experiment

    SciTech Connect

    Strain, R.V.; Konicek, C.F.; Tsai, H.

    1996-10-01

    Biaxial creep specimens will be included in the HFIR RB-12J experiment to study in-reactor creep of the V-4Cr-4Ti alloy at {approx}500{degrees}C and 5 dpa. These specimens were fabricated with the 500-kg, heat (832665) material and pressurized to attain 0, 50, 100, 150, and 200 MPa mid-wall hoop stresses during the irradiation.

  11. Wave-packet treatment of reactor neutrino oscillation experiments and its implications on determining the neutrino mass hierarchy

    NASA Astrophysics Data System (ADS)

    Chan, Yat-Long; Chu, M.-C.; Tsui, Ka Ming; Wong, Chan Fai; Xu, Jianyi

    2016-06-01

    We derive the neutrino flavor transition probabilities with the neutrino treated as a wave packet. The decoherence and dispersion effects from the wave-packet treatment show up as damping and phase-shifting of the plane-wave neutrino oscillation patterns. If the energy uncertainty in the initial neutrino wave packet is larger than around 0.01 of the neutrino energy, the decoherence and dispersion effects would degrade the sensitivity of reactor neutrino experiments to mass hierarchy measurement to lower than 3 σ confidence level.

  12. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  13. Some parameter boundaries governing microgravity pool boiling modes.

    PubMed

    Merte, Herman

    2006-09-01

    Pool boiling experiments were conducted in microgravity on five space shuttle flights, using a flat plate heater consisting of a semitransparent thin gold film deposited on a quartz substrate that also acted as a resistance thermometer. The test fluid was R-113, and the vapor bubble behavior at the heater surface was photographed from beneath as well as from the side. Each flight consisted of a matrix of three levels of imposed heat flux and three levels of initial bulk liquid subcooling. In many of the total of 45 experiments, steady nucleate boiling was observed from 16-mm movie films, where a large vapor bubble formed and remained slightly removed from the heater surface, with small vapor bubbles growing on the heater surface, and on contact coalescing with the large bubble. Computations of the forces associated with the momentum transfer in this process, which counters the Marangoni convection effects tending to impel the large bubble toward the heater surface, have been completed for all cases where applicable. The modes of pool boiling observed with successive increases in levels of heat flux in microgravity are categorized as: (i) minimum or incipient nucleate boiling; (ii) nucleate boiling with vigorous motion of the bubbles adjacent and parallel to the heater surface, impelled by Marangoni convection effects; (iii) nucleate boiling followed by coalescence with a neighboring large vapor bubble; (iv) partial dryout of the heater surface, in parallel with nucleate boiling; (v) complete dryout. The boundaries between these modes are delineated graphically as a function of the imposed heat flux and initial bulk liquid subcooling, together with the levels of the forces holding the large bubbles, acting as vapor reservoirs, away from the heater surface for the steady nucleate boiling mode. PMID:17124149

  14. Bandages of boiled potato peels.

    PubMed

    Patil, A R; Keswani, M H

    1985-08-01

    The use of potato peels as a dressing for burn wounds has been reported previously. A technique of preparing bandage rolls with boiled potato peels is now presented, which makes dressing of a burn wound more convenient. PMID:4041947

  15. The coolability limits of a reactor pressure vessel lower head

    SciTech Connect

    Theofanous, T.G.; Syri, S.

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  16. Efficient design of experiments for complex response surfaces with application to etching uniformity in a plasma reactor

    NASA Astrophysics Data System (ADS)

    Tatavalli Mittadar, Nirmal

    Plasma etching uniformity across silicon wafers is of paramount importance in the semiconductor industry. The complexity of plasma etching, coupled with lack of instrumentation to provide real-time process information (that could be used for feedback control), necessitate that optimal conditions for uniform etching must be designed into the reactor and process recipe. This is often done empirically using standard design of experiments which, however, are very costly and time consuming. The objective of this study was to develop a general purpose efficient design strategy that requires a minimum number of experiments, and can handle complex constraints in the presence of uncertainties. Traditionally, Response Surface Methodology (RSM) is used in these applications to design experiments to determine the optimal value of decision variables or inputs. We demonstrated that standard RSM, when applied to the problem of plasma etching uniformity, has the following drawbacks (1) inefficient search due to process nonlinearities, (2) lack of converge to the optimum, and, (3) inability to handle complex inequality constraints. We developed a four-phase Efficient Design Strategy (EDS) based on the DACE paradigm (Design and Analysis of Computer Experiments) and Bayesian search algorithms. The four phases of EDS are: (1) exploration of the design space by maximizing information, (2) exploration of the design space for feasible points by maximizing probability of constraint satisfaction, (3) optimization of the objective and (4) constrained local search. We also designed novel algorithms to switch between the different phases. The choice of model parameters for DACE predictors is usually determined by the Maximum Likelihood Estimation (MLE) method. Depending on the dataset, MLE could result in unrealistic predictors that show a peak-and-dip behavior. To solve this problem we developed techniques to detect the presence of peak-and-dip behavior and a new scheme based on Maximum a

  17. Waterproofed photomultiplier tube assemblies for the Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    Chow, Ken; Cummings, John; Edwards, Emily; Edwards, William; Ely, Ry; Hoff, Matthew; Lebanowski, Logan; Li, Bo; Li, Piyi; Lin, Shih-Kai; Liu, Dawei; Liu, Jinchang; Luk, Kam-Biu; Miao, Jiayuan; Napolitano, Jim; Ochoa-Ricoux, Juan Pedro; Peng, Jen-Chieh; Qi, Ming; Steiner, Herbert; Stoler, Paul; Stuart, Mary; Wang, Lingyu; Yang, Changgen; Zhong, Weili

    2015-09-01

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

  18. Flow boiling in vertical down-flow

    SciTech Connect

    Dougherty, T.; Fighetti, C.; Reddy, G.; Yang, B.; Jafri, T. ); McAssey, E. ); Qureshi, Z. )

    1989-01-01

    An experimental program has been conducted to investigate the onset of Ledinegg instability in vertical down-flow. For three size uniformly heated test sections with L/D ratios from 100 to 150, the pressure drop under subcooled boiling conditions has been obtained for a wide range of operating parameters. The results are presented in non-dimensional forms which correlate the important variables and provide techniques for predicting the onset of flow instability. 3 refs.

  19. Flow boiling in vertical down-flow

    SciTech Connect

    Dougherty, T.; Fighetti, C.; Reddy, G.; Yang, B.; Jafri, T.; McAssey, E.; Qureshi, Z.

    1989-12-31

    An experimental program has been conducted to investigate the onset of Ledinegg instability in vertical down-flow. For three size uniformly heated test sections with L/D ratios from 100 to 150, the pressure drop under subcooled boiling conditions has been obtained for a wide range of operating parameters. The results are presented in non-dimensional forms which correlate the important variables and provide techniques for predicting the onset of flow instability. 3 refs.

  20. Bubble transport in subcooled flow boiling

    NASA Astrophysics Data System (ADS)

    Owoeye, Eyitayo James

    Understanding the behavior of bubbles in subcooled flow boiling is important for optimum design and safety in several industrial applications. Bubble dynamics involve a complex combination of multiphase flow, heat transfer, and turbulence. When a vapor bubble is nucleated on a vertical heated wall, it typically slides and grows along the wall until it detaches into the bulk liquid. The bubble transfers heat from the wall into the subcooled liquid during this process. Effective control of this transport phenomenon is important for nuclear reactor cooling and requires the study of interfacial heat and mass transfer in a turbulent flow. Three approaches are commonly used in computational analysis of two-phase flow: Eulerian-Lagrangian, Eulerian-Eulerian, and interface tracking methods. The Eulerian- Lagrangian model assumes a spherical non-deformable bubble in a homogeneous domain. The Eulerian-Eulerian model solves separate conservation equations for each phase using averaging and closure laws. The interface tracking method solves a single set of conservation equations with the interfacial properties computed from the properties of both phases. It is less computationally expensive and does not require empirical relations at the fluid interface. Among the most established interface tracking techniques is the volume-of-fluid (VOF) method. VOF is accurate, conserves mass, captures topology changes, and permits sharp interfaces. This work involves the behavior of vapor bubbles in upward subcooled flow boiling. Both laminar and turbulent flow conditions are considered with corresponding pipe Reynolds number of 0 -- 410,000 using a large eddy simulation (LES) turbulence model and VOF interface tracking method. The study was performed at operating conditions that cover those of boiling water reactors (BWR) and pressurized water reactors (PWR). The analysis focused on the life cycle of vapor bubble after departing from its nucleation site, i.e. growth, slide, lift-off, rise