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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. Stability monitor experience in German BWRs (boiling water reactor)

    SciTech Connect

    Goldstein, L. ); Fuge, R. ); Seepolt, R. ); Frank, M. )

    1989-11-01

    A digital stability monitor developed by NIS Ingenieurgesellschaft MBH, operable on a personal computer, is in use at three boiling water reactor (BWR) plants in the Federal Republic of Germany (FRG). The device has received the approval of an FRG licensing authority. It has been in operation for 5 yr. The stability monitor is a measurement device used to accurately determine and confirm the boundaries of the exclusion region on the operating power-flow map, and to permit controlled operation through such regions. This is achieved via neutron noise measurements converted to direct readout of decay ratios. It has satisfied regulatory requirements by defining potential unstable operating regions at Kernkraftwerk Isar in the FRG and the high power density Gundremmingen (KRB Units B and C) BWRs.

  4. 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.

  5. 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.

  6. 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)

  7. Design study of an irradiation experiment with inert matrix and mixed-oxide fuel at the Halden boiling water reactor

    NASA Astrophysics Data System (ADS)

    Kasemeyer, U.; Joo, H.-K.; Ledergerber, G.

    1999-08-01

    An effective way to reduce the large quantities of plutonium currently accumulated worldwide would be to use uranium-free fuel in light water reactors (LWRs) so that no new plutonium is produced. To test such a new fuel under reactor conditions and in comparison with standard mixed-oxide (MOX) fuel, an irradiation experiment is planned at the Halden boiling water reactor. The behaviour of three fuel rods consisting of uranium-free fuel will be investigated together with three rods made out of uranium-plutonium mixed-oxide fuel in the same assembly. The fuel compositions were adjusted so that all rods produce a similar power. Because of the moderation with D 2O in the Halden reactor, two different surroundings of the considered assembly were examined to analyze the influence of the flux spectrum on the experiment. This showed that the influence of the spectrum on the material behaviour is negligible. The relation between assembly power and average neutron detector signal as well as the burnup or depletion function was calculated. The assumed power history was adapted to a usual LWR schedule. It is possible to reach a burnup of ˜540 MW d kg HM-1 with the uranium-free fuel and ˜54 MW d kg HM-1 with the MOX fuel after five years of irradiation, which is similar to the average burnup reached in commercial LWRs after four years of operation.

  8. 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.

  9. (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.

  10. 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.

  11. SWR 1000: The Innovative Boiling Water Reactor

    SciTech Connect

    Brettschuh, Werner; Hudson, Greg

    2004-07-01

    Framatome ANP has developed the boiling water reactor SWR 1000 in close cooperation with German nuclear utilities and with support from various European partners. This advanced reactor design marks a new era in the successful tradition of boiling water reactor technology and, with a gross electric output of between 1290 and 1330 MW, is aimed at assuring competitive power generating costs compared to gas- and coal-fired stations. At the same time, the SWR 1000 meets the highest safety standards, including control of a core melt accident these objectives are met by supplementing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. The SWR 1000 fulfills international nuclear regulatory requirements and has been offered to TVO for the fifth nuclear unit in Finland. (authors)

  12. 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)

  13. Boiling water reactor licensing basis transient

    SciTech Connect

    Cheng, H. S.; Lu, M. S.; Shier, W. G.; Diamond, D. J.; Levine, M. M.; Odar, F.

    1980-01-01

    An analysis is presented of the licensing basis transient for a boiling water reactor where a turbine trip occurs without steam bypass. The analysis was performed by means of the two-dimensional (R,Z) core dynamics code BNL-TWIGL in conjunction with the system transient code RELAP-3B. Two plant models were used and produced similar results for the analysis of the Peach Bottom turbine trip tests. The models differed in the representation of the steam separator. The analysis of the licensing basis transient produced somewhat different results. The results of sensitivity studies to help explain the differences are presented as well as an analysis of the licensing basis transient with recirculation pump trip. 2 refs., 17 figs., 1 tab.

  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. NUCLEAR SUPERHEATER FOR BOILING WATER REACTOR

    DOEpatents

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

    1962-05-15

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

  16. Flow Boiling and Condensation Experiment

    NASA Image and Video Library

    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...

  17. 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.

  18. 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

    ... 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 18... 3, 2010. Antonio F. Dias, Chief, Reactor Safety Branch B, Advisory Committee on Reactor...

  19. Chemical Characterization of Simulated Boiling Water Reactor Coolant

    DTIC Science & Technology

    1990-05-01

    industry to reduce personnel radiation exposure and down-time associated with the operation, mainte- nance and refueling of Light Water Reactor (LWR...AD-A226 654 t t-FILL UIY C CHEMICAL CHARACTERIZATION OF SIMULATED , .BOILING WATER REACTOR COOLANt by Li . . , . , - VERRDON HOLBROOK MASON f ; B.S...CHARACTERIZATION OF SIMULATED BOILING WATER REACTOR COOLANT by VERRDON HOLBROOK MASON Submitted to the Department of Nuclear Engineering on May 9, 1988 in

  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. 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

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

  2. Experiment for search for sterile neutrino at SM-3 reactor

    NASA Astrophysics Data System (ADS)

    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.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2016-11-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

  3. A New Computational Tool for Simulation of 3-D Flow and Heat Transfer in Boiling Water Reactors

    SciTech Connect

    Chen, Hudong

    2002-12-09

    This Phase I work has developed a novel hybrid Lattice Boltzmann Model for the simulation of nonideal fluid thermal dynamics and demonstrated that this model can be used to simulate fundamental two-phase flow processes including boiling initiation, bubble formation and coalescency, and flow-regime formation.

  4. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 3-flavor oscillations with current and future reactor experiments

    NASA Astrophysics Data System (ADS)

    Dwyer, Dan

    2017-01-01

    Nuclear reactors have been a crucial tool for our understanding of neutrinos. The disappearance of electron antineutrinos emitted by nuclear reactors has firmly established that neutrino flavor oscillates, and that neutrinos consequently have mass. The current generation of precision measurements rely on some of the world's most intense reactor facilities to demonstrate that the electron antineutrino mixes with the third antineutrino mass eigenstate (v3-). Accurate measurements of antineutrino energies robustly determine the tiny difference between the masses-squared of the v3- state and the two more closely-spaced v1- and v2- states. These results have given us a much clearer picture of neutrino mass and mixing, yet at the same time open major questions about how to account for these small but non-zero masses in or beyond the Standard Model. These observations have also opened the door for a new generation of experiments which aim to measure the ordering of neutrino masses and search for potential violation of CP symmetry by neutrinos. I will provide a brief overview of this exciting field. Work supported under DOE OHEP DE-AC02-05CH11231.

  10. 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... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Chemical Gradients in Crud on Boiling Water Reactor Fuel Elements

    SciTech Connect

    D. L. Porter; D. E. Janney

    2007-04-01

    Crud (radioactive corrosion products formed inside nuclear reactors is a major problem in commercial power-producing nuclear reactors. Although there are numerous studies of simulated (non-radioactive) crud, characteristics of crud from actual reactors are rarely studied. This study reports scanning electron microscope (SEM) studies of fragments of crud from a commercially operating boiling water reactor. Chemical analyses in the SEM indicated that the crud closest to the outer surfaces of the fuel pins in some areas had Fe:Zn ratios close to 2:1, which decreased away from the fuel pin in some of the fragments. In combination with transmission electron microsope analyses (published elsewhere), these results suggest that the innermost layer of crud in some areas may consist of franklinite (ZnFe2O4, also called zinc spinel), while outer layers in these areas may be predominantly iron oxides.

  17. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

  18. 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)

  19. Electrically Driven Liquid Film Boiling Experiment

    NASA Technical Reports Server (NTRS)

    Didion, Jeffrey R.

    2016-01-01

    This presentation presents the science background and ground based results that form the basis of the Electrically Driven Liquid Film Boiling Experiment. This is an ISS experiment that is manifested for 2021. Objective: Characterize the effects of gravity on the interaction of electric and flow fields in the presence of phase change specifically pertaining to: a) The effects of microgravity on the electrically generated two-phase flow. b) The effects of microgravity on electrically driven liquid film boiling (includes extreme heat fluxes). Electro-wetting of the boiling section will repel the bubbles away from the heated surface in microgravity environment. Relevance/Impact: Provides phenomenological foundation for the development of electric field based two-phase thermal management systems leveraging EHD, permitting optimization of heat transfer surface area to volume ratios as well as achievement of high heat transfer coefficients thus resulting in system mass and volume savings. EHD replaces buoyancy or flow driven bubble removal from heated surface. Development Approach: Conduct preliminary experiments in low gravity and ground-based facilities to refine technique and obtain preliminary data for model development. ISS environment required to characterize electro-wetting effect on nucleate boiling and CHF in the absence of gravity. Will operate in the FIR - designed for autonomous operation.

  20. 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.

  1. Hydrogen Water Chemistry Technology in Boiling Water Reactors

    SciTech Connect

    Lin, Chien C

    2000-04-15

    Modification of coolant chemistry by feedwater hydrogen addition in boiling water reactors (BWRs), generally called hydrogen water chemistry (HWC), is a viable option to mitigate the intergranular stress corrosion cracking problems for operating BWRs. Some fundamentals in HWC technologies as known today are reviewed. Several full-scale HWC test results are analyzed, and the roles that hydrogen plays in HWC technology are identified and quantitatively evaluated. Some deficiencies in water radiolysis modeling for BWR applications under HWC conditions and the impact of {sup 16}N radiation field increase in the main steam line are also discussed.

  2. 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.

  3. 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)

  4. Analysis of scrams and forced outages at boiling water reactors

    SciTech Connect

    Earle, R. T.; Sullivan, W. P.; Miller, K. R.; Schwegman, W. J.

    1980-07-01

    This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability.

  5. 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)

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

    SciTech Connect

    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.

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

    DOE PAGES

    Vo, D.; Favalli, A.; Grogan, B.; ...

    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 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

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

    SciTech Connect

    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.

  9. 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.

  10. 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.

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

    SciTech Connect

    Baversten, B.; Linden, M.J.

    1995-03-01

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

  12. 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

    ... 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 11545...: February 12, 2010. Antonio F. Dias, Chief Reactor Safety Branch B, Advisory Committee on Reactor...

  13. 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)

  14. RELAP5/MOD3 subcooled boiling model assessment

    SciTech Connect

    Devkin, A.S.; Podosenov, A.S.

    1998-05-01

    This report presents the assessment of the RELAP5/Mod3 (5m5 version) code subcooled boiling process model which is based on a variety of experiments. The accuracy of the model is confirmed for a wide range of regime parameters for the case of uniform heating along the channel. The condensation rate is rather underpredicted, which may lead to considerable errors in void fraction behavior prediction in subcooled boiling regimes for nonuniformly or unheated channels.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Radial nodalization effects on BWR (boiling water reactor) stability calculations

    SciTech Connect

    March-Leuba, J.

    1990-01-01

    Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters and modeling assumptions. In particular, the number of thermohydraulic regions (i.e., channels) used in the calculation can affect the results of decay ratio calculations by as much as 30%. This paper presents the background theory behind the observed effects of radial nodalization in BWR stability calculations. The theory of how a radial power distribution can be simulated in time or frequency domain codes by using representative'' regions is developed. The approximations involved in this method of solution are reviewed, and some examples of the effect of radial nodalization are presented based on LAPUR code solutions. 2 refs., 4 figs., 2 tabs.

  20. 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.

  1. 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 ...

  2. 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.

  3. 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.

  4. On Stability of Natural-circulation-cooled Boiling Water Reactors during Start-up (Experimental Results)

    SciTech Connect

    Manera, A.; Van der Hagen, T.H.J.J.

    2002-07-01

    The characteristics of flashing-induced instabilities, which are of importance during the start-up phase of natural-circulation Boiling Water Reactors (BWRs), are studied. Experiments at typical start-up conditions (low power and low pressure) are carried out on a steam/water natural circulation loop. The mechanism of flashing-induced instability is analyzed in detail and it is found that non-equilibrium between phases and enthalpy transport plays an important role in the instability process. Pressure and steam volume in the steam dome are found to have a stabilizing effect. The main characteristics of the instabilities have been analyzed. (authors)

  5. 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

  6. 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.

  7. 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)

  8. Boils

    MedlinePlus

    ... or recurrent boils, which are usually due to Staph infections. The bacteria are picked up somewhere and then ... version of boils is folliculitis . This is an infection of hair follicles, usually with Staph bacteria. These often itch more than hurt. The ...

  9. Nondestructive assay of spent boiling water reactor fuel by active neutron interrogation

    SciTech Connect

    Blakeman, E.D.; Ricker, C.W.; Ragan, G.L.; Difilippo, F.C.; Slaughter, G.G.

    1981-01-01

    Spent boiling water reactor (BWR) fuel from Dresden I was assayed for total fissile mass, using the active neutron interrogation method. The nondestructive assay (NDA) system used has four Sb-Be sources for interrogation of the fuels; the induced fission neutrons from the fuel are counted by four lead-shielded methane-filled proportional counters biased above the energy of the source neutrons. Spent fuel rods containing 9 kg of heavy metal were chopped into 5-cm segments and loaded into three 1-liter cans. The three cans were assayed in seven combinations of one, two, or three cans, enabling an evaluation of the precision and accuracy of the NDA system for different amounts of fissile material. The fissile mass in each combination was determined by comparing the induced-fission-neutron counts with the counts obtained from a known standard comprising chopped segments of unirradiated Dresden fuel. These masses were compared to the masses determined by chemical analyses of the spent fuel. The results from the nondestructive assays agreed with results from the chemical analyses to within 2 to 3%. Similar agreement was obtained when two combinations of canned spent fuel were used as standards for the nondesctuctive assays. The assay of BWR spent fuel served as a test of the NDA system which was developed at the Oak Ridge National Laboratory for the assay of spent liquid metal fast breeder reactor (LMFBR) fuel subassemblies at the heat-end of a reprocessing plant. Results of previous experiments and calculations reported earlier using simulated LMFBR fuel subassemblies indicated that the NDA system can measure the fissile masses of spent fuel subassemblies to within an accuracy of 3%. Results of the assays of spent BWR fuel reported herein support this conclusion.

  10. Nondestructive assay of spent boiling-water-reactor fuel by active neutron interrogation

    SciTech Connect

    Blakeman, E.D.; Ricker, C.W.; Ragan, G.L.; Difilippo, F.C.; Slaughter, G.G.

    1981-01-01

    Spent boiling water reactor (BWR) fuel from Dresden I was assayed for total fissile mass, using the active neutron interrogation method. The nondestructive assay (NDA) system used has four Sb-Be sources for interrogation of the fuels; the induced fission neutrons from the fuel are counted by four lead-shielded methane-filled proportional counters biased above the energy of the source neutrons. Results agreed with results from the chemical analyses to within 2 to 3%. Similar agreement was obtained when two combinations of canned spent fuel were used as standards for the nondestructive assays.

  11. 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.

  12. 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.

  13. N Reactor departure from nucleate boiling correlation for outer annulus subchannel at higher steam quality

    SciTech Connect

    Wittekind, W.D.

    1988-02-01

    This report justifies an N Reactor departure from nucleate boiling heat flux correlation for the outer annulus subchannel into the higher steam quality region. The center hole flow channel was completed in 1963, and the inner annulus flow channel was completed in 1978. This extension to departure from nucleate boiling correlation will complete the extension into the higher steam quality region and relieve some calculated thermal hydraulic limits without compromising reactor safety. Justification for this departure from nucleate boiling heat flux correlation extension into the higher steam quality region results from the following references: Annular flow channel correlations, Annular flow channel measurements, and Hanford Site measurements on an N Reactor model. This extension to departure from nuclear boiling heat flux correlation is reasonable and conservative. 19 refs., 4 figs., 14 tabs.

  14. Nucleate boiling pressure drop in an annulus: Book 3

    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.

  15. (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.

  16. 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

    ... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

  17. 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

    ... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

  18. 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.

  19. Pulsational characteristics of the natural-circulation loop of a large-scale model of a light-boiling boiling-water reactor

    SciTech Connect

    Babykin, A.S.; Balunov, B.F.; Chernykh, N.G.; Smirnov, E.L.; Tisheninova, V.I.; Zhiuitskaya, T.S.

    1985-10-01

    The results of an experimental study of a natural-circulation (NC) loop, whose geometrical and hydraulic characteristics are presented are described. The range of state parameters encompassed in the experiments is also indicated. The authors used a large-scale model of a low-boiling-water reactor, with natural heights and reduced stages of separate elements of the NC loop. The study confirmed that under the conditions the pulsations in the flow rate of the coolant occurs only in the transitional zone from natural circulation of the singlephase medium to natural circulation of the two-phase coolant.

  20. Packed Bed Reactor Experiment

    NASA Image and Video Library

    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 ...

  1. 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

  2. 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)

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Physical characteristics of GE (General Electric) BWR (boiling-water reactor) fuel assemblies

    SciTech Connect

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs.

  10. 77 FR 16098 - In the Matter of All Operating Boiling Water Reactor Licensees With Mark I and Mark II...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-19

    ... Communications.'' The Director, Office of Nuclear Reactor Regulation may, in writing, relax or rescind any of the... writing to the Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission... Nuclear Reactor Regulation. Operating Boiling Water Reactor Licenses With Mark I and Mark II...

  11. Creation of neutrino laboratory for carrying out experiment on search for a sterile neutrino at the SM-3 reactor

    NASA Astrophysics Data System (ADS)

    Serebrov, A. P.; Ivochkin, V. G.; Samoilov, R. M.; Fomin, A. K.; Zinov'ev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinskii, N. V.; Solovei, V. A.; Chernyi, A. V.; Zherebtsov, O. M.; Martem'yanov, 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.; Gromov, M. O.; Afanas'ev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2015-12-01

    To check the existence of a sterile neutrino, a neutrino laboratory aimed at searching reactor antineutrino oscillations is created at the SM-3 reactor. A prototype of a neutrino detector with a scintillator volume of 400 L is moved at distances 6-11 m from the core of the reactor. Background conditions are measured. It is shown that the cosmic rays background is the main problem in the experiment. The prospects of the search for reactor antineutrino oscillations at short distances are discussed.

  12. Characterization of phases in ‘crud’ from boiling-water reactors by transmission electron microscopy

    NASA Astrophysics Data System (ADS)

    Janney, Dawn E.; Porter, Douglas L.

    2007-05-01

    This paper reports phases identified in samples of crud (activated corrosion products) from two commercial boiling-water reactors using transmission and analytical electron microscopy and selected-area electron diffraction. Franklinite (ZnFe 2O 4) was observed in both samples. Hematite (α-Fe 2O 3), crystalline silica (SiO 2), a fine-grained mixture of iron oxides probably including magnetite (Fe 3O 4), hematite (α-Fe 2O 3), and goethite (α-FeOOH), and an unidentified high-Ba, high-S phase were observed in one of the samples. Willemite (Zn 2SiO 4), amorphous silica, and an unidentified iron-chromium phase were observed in the other. Chloride-bearing phases were found in both samples, and are assumed to represent sample contaminants. Because of the small sample volumes and numbers of particles studied and the possibility of contamination, it is not clear whether the differences between the phases observed in the two crud samples represent actual differences in the assemblages formed in the reactors.

  13. 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

  14. REFLECTOR CONTROL OF A BOILING-WATER REACTOR

    DOEpatents

    Treshow, M.

    1962-05-22

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

  15. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    SciTech Connect

    Aksan, S.N.; Stierli, F.; Analytis, G.T.

    1992-03-01

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended.

  16. A microgravity boiling and convective condensation experiment

    NASA Technical Reports Server (NTRS)

    Kachnik, Leo; Lee, Doojeong; Best, Frederick; Faget, Nanette

    1987-01-01

    A boiling and condensing test article consisting of two straight tube boilers, one quartz and one stainless steel, and two 1.5 m long glass-in-glass heat exchangers, on 6 mm ID and one 10 mm ID, was flown on the NASA KC-135 0-G aircraft. Using water as the working fluid, the 5 kw boiler produces two phase mixtures of varying quality for mass flow rates between 0.005 and 0.1 kg/sec. The test section is instrumented at eight locations with absolute and differential pressure transducers and thermocouples. A gamma densitometer is used to measure void fraction, and high speed photography records the flow regimes. A three axis accelerometer provides aircraft acceleration data (+ or - 0.01G). Data are collected via an analog-to-digital conversion and data acquisition system. Bubbly, annular, and slug flow regimes were observed in the test section under microgravity conditions. Flow oscillations were observed for some operating conditions and the effect of the 2-G pullout prior to the 0-G period was observed by continuously recording data throughout the parabolas. A total fo 300 parabolas was flown.

  17. A microgravity boiling and convective condensation experiment

    NASA Astrophysics Data System (ADS)

    Kachnik, Leo; Lee, Doojeong; Best, Frederick; Faget, Nanette

    1987-12-01

    A boiling and condensing test article consisting of two straight tube boilers, one quartz and one stainless steel, and two 1.5 m long glass-in-glass heat exchangers, on 6 mm ID and one 10 mm ID, was flown on the NASA KC-135 0-G aircraft. Using water as the working fluid, the 5 kw boiler produces two phase mixtures of varying quality for mass flow rates between 0.005 and 0.1 kg/sec. The test section is instrumented at eight locations with absolute and differential pressure transducers and thermocouples. A gamma densitometer is used to measure void fraction, and high speed photography records the flow regimes. A three axis accelerometer provides aircraft acceleration data (+ or - 0.01G). Data are collected via an analog-to-digital conversion and data acquisition system. Bubbly, annular, and slug flow regimes were observed in the test section under microgravity conditions. Flow oscillations were observed for some operating conditions and the effect of the 2-G pullout prior to the 0-G period was observed by continuously recording data throughout the parabolas. A total fo 300 parabolas was flown.

  18. 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

  19. Formation and deposition of platinum nanoparticles under boiling water reactor conditions

    NASA Astrophysics Data System (ADS)

    Grundler, Pascal V.; Veleva, Lyubomira; Ritter, Stefan

    2017-10-01

    Stress corrosion cracking (SCC) is a well-known degradation mechanism for components of boiling water reactors (BWRs). Therefore the mitigation of SCC is important for ensuring the integrity of the reactor system. Noble metal chemical application (NMCA) has been developed by General Electric to mitigate SCC and reduce the negative side-effects of hydrogen water chemistry used initially for SCC mitigation. NMCA is now widely applied as an online process (OLNC) during power operation. However, the understanding of the parameters that control the formation and deposition of the noble metal (Pt) particles in a BWR was still incomplete. To fill this knowledge gap, systematic studies on the formation and deposition behaviour of Pt particles in simulated and real BWR environment were performed in the framework of a research project at PSI. The present paper summarizes the most important findings. Experiments in a sophisticated high-temperature water loop revealed that the flow conditions, water chemistry, the Pt injection rate, and the pre-conditioning of the stainless steel surfaces have an impact on the Pt deposition behaviour. Slower Pt injection rates and stoichiometric excess of H2 over O2 produce smaller particles, which may increase the efficiency of the OLNC technique in mitigating SCC. Surfaces with a well-developed oxide layer retain more Pt particles. Furthermore, the pre- and post-OLNC exposure times play an important role for the Pt deposition on specimens exposed at the KKL power plant. Redistribution of Pt in the plant takes place, but most of the Pt apparently does not redeposit on the steel surfaces in the reactor system. Comparison of lab and plant results also demonstrated that plant OLNC applications can be simulated reasonably well on the lab scale.

  20. 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.

  1. Nucleate Pool Boiling Experiments (NPBX) on the International Space Station

    NASA Astrophysics Data System (ADS)

    Dhir, Vijay Kumar; Warrier, Gopinath R.; Aktinol, Eduardo; Chao, David; Eggers, Jeffery; Sheredy, William; Booth, Wendell

    2012-11-01

    During the period of March-May 2011, a series of boiling experiments was carried out in the Boiling Experimental Facility (BXF) located in the Microgravity Science Glovebox (MSG) of the International Space Station (ISS). The BXF Facility was carried to ISS on Space Shuttle Mission STS-133 on February 24, 2011. Nucleate Pool Boiling Experiment (NPBX) was one of the two experiments housed in the BXF. Results of experiments on single bubble dynamics (e.g., inception and growth), multiple bubble dynamics (lateral merger and departure, if any), nucleate pool boiling heat transfer, and critical heat flux are described. In the experiments Perfluoro-n-hexane was used as the test liquid. The system pressure was varied from 51 to 243 kPa, pool temperature was varied from 30° to 59°C, and test surface temperature was varied from 40° to 80°C. The test surface was a polished aluminum disc (1 mm thick, 89.5 mm in diameter) heated from below with strain gage heaters. Five cylindrical cavities were formed on the surface with four cavities located at the corners of a square and one in the middle. During experiments the magnitude of mean gravity level normal to the heater surface varied from 1.2 × 10 - 7g e to 6 × 10 - 7g e . The results of the experiments show that a single bubble continues to grow to occupy the size of the chamber without departing from the heater surface. During lateral merger of bubbles, at high superheats a large bubble may lift off from the surface but continues to hover near the surface. Neighboring bubbles are continuously pulled into the large bubble. At low superheats bubbles at neighboring sites simply merge to yield a larger bubble. The larger bubble mostly locates in the middle of the heated surface and serves as a vapor sink. The latter mode continues to persist when boiling is occurring all over the heater surface. Heat fluxes for steady state nucleate boiling and critical heat fluxes are found to be much lower than those obtained under earth

  2. 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.

  3. 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)

  4. Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors

    SciTech Connect

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).

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

    SciTech Connect

    Beyer, Carl E.; Geelhood, Kenneth J.

    2013-06-01

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

  6. Local stability tests in Dresden 2 boiling water reactor

    SciTech Connect

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

  7. The effects of aging on Boiling Water Reactor core isolation cooling system

    SciTech Connect

    Lee, Bom Soon

    1994-06-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes.

  8. 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.

  9. Time domain model sensitivity in boiling water reactor stability analysis using TRAC/BF1

    SciTech Connect

    Borkowski, J.A. ); Robinson, G.E.; Baratta, A.J.; Kattic, M. . Dept. of Nuclear Engineering)

    1993-07-01

    Boiling water nuclear reactors (BWRs) may experience density wave instabilities. These instabilities cause the density, and consequently the mass flow rate, to oscillate in the shrouded fuel bundles. This effect causes the nuclear power generation to oscillate because of the tight coupling of flow to power, especially under gravity-driven circulation. To predict the amplitude of the power oscillation, a time domain transient analysis tool may be employed. The modeling tool must have sufficient hydrodynamic detail to model natural circulation in two-phase flow as well as the coupled nuclear feedback. TRAC/BF1 is a modeling code with such capabilities. A dynamic system model is developed for a typical BWR. Using this tool, it is demonstrated that density waves may be modeled in this fashion and that their resultant hydrodynamic and nuclear behavior correspond well to simple theory. Several cases are analyzed using this model, the goal being to determine the coupling between the channel hydrodynamics and the nuclear power. As predicted by others, the two-phase friction controls the extent of the oscillation. Because of this sensitivity, existing conventional methodologies of implementing two-phase friction into analysis codes of this type can lead to significant deviation in results from one case to another. It is found that higher dimensional nuclear feedback models reduce the extent of the oscillation.

  10. 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.

  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. A pilot reactor study on the effect of the naphtha boiling point properties in catalytic reforming

    SciTech Connect

    Moliord, K.; Tanem, I.; Grande, K.

    1995-12-31

    Three naphthas with different initial and three naphthas with different final boiling points were compared by testing in a pilot reactor. The pilot reactor unit consisted of isothermal, once-through 200 cm{sup 2} reactors with on-line GCs for full product analysis and octane number determination. Octane numbers, reformate yields and composition, gas and hydrogen yields were measured as function of reaction temperature at 16 bar reaction pressure and a molar H{sub 2}/HC ratio of 4.23. Catalyst deactivation was studied over 2 weeks periods at high seventy conditions, i.e. 102.4 RON and a H{sub 2}/HC ratio of 2.2. Test results, with emphasis on the yields of benzene and other aromatics, hydrogen yields as well as catalyst deactivation, are presented.

  13. 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.

  14. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    SciTech Connect

    Nielsen, Joseph; Tokuhiro, Akira; Hiromoto, Robert; Tu, Lei

    2015-11-13

    Evaluation of the impacts of uncertainty and sensitivity in modeling presents a significant set of challenges in particular to high fidelity modeling. Computational costs and validation of models creates a need for cost effective decision making with regards to experiment design. Experiments designed to validate computation models can be used to reduce uncertainty in the physical model. In some cases, large uncertainty in a particular aspect of the model may or may not have a large impact on the final results. For example, modeling of a relief valve may result in large uncertainty, however, the actual effects on final peak clad temperature in a reactor transient may be small and the large uncertainty with respect to valve modeling may be considered acceptable. Additionally, the ability to determine the adequacy of a model and the validation supporting it should be considered within a risk informed framework. Low fidelity modeling with large uncertainty may be considered adequate if the uncertainty is considered acceptable with respect to risk. In other words, models that are used to evaluate the probability of failure should be evaluated more rigorously with the intent of increasing safety margin. Probabilistic risk assessment (PRA) techniques have traditionally been used to identify accident conditions and transients. Traditional classical event tree methods utilize analysts’ knowledge and experience to identify the important timing of events in coordination with thermal-hydraulic modeling. These methods lack the capability to evaluate complex dynamic systems. In these systems, time and energy scales associated with transient events may vary as a function of transition times and energies to arrive at a different physical state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This study presents a methodology to address

  15. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    DOE PAGES

    Nielsen, Joseph; Tokuhiro, Akira; Hiromoto, Robert; ...

    2015-11-13

    Evaluation of the impacts of uncertainty and sensitivity in modeling presents a significant set of challenges in particular to high fidelity modeling. Computational costs and validation of models creates a need for cost effective decision making with regards to experiment design. Experiments designed to validate computation models can be used to reduce uncertainty in the physical model. In some cases, large uncertainty in a particular aspect of the model may or may not have a large impact on the final results. For example, modeling of a relief valve may result in large uncertainty, however, the actual effects on final peakmore » clad temperature in a reactor transient may be small and the large uncertainty with respect to valve modeling may be considered acceptable. Additionally, the ability to determine the adequacy of a model and the validation supporting it should be considered within a risk informed framework. Low fidelity modeling with large uncertainty may be considered adequate if the uncertainty is considered acceptable with respect to risk. In other words, models that are used to evaluate the probability of failure should be evaluated more rigorously with the intent of increasing safety margin. Probabilistic risk assessment (PRA) techniques have traditionally been used to identify accident conditions and transients. Traditional classical event tree methods utilize analysts’ knowledge and experience to identify the important timing of events in coordination with thermal-hydraulic modeling. These methods lack the capability to evaluate complex dynamic systems. In these systems, time and energy scales associated with transient events may vary as a function of transition times and energies to arrive at a different physical state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This study presents a methodology to address

  16. 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.

  17. The improvement of the interfacial drag model in RELAP5/MOD3.3 to simulate downcomer boiling phenomena in APR1400

    SciTech Connect

    Kim, Han-Gon; Lee, Seok-Ho

    2006-07-01

    In late reflood phase of LBLOCA, the injected water flow-rate is small compared to those in refill and early reflood phases due to the termination of large cooling water source, that is, the Safety Injection Tanks (SITs). At this situation, the water in downcomer could be vaporized near the reactor vessel wall surface having stored energy. The technical issue is if this local boiling could be extended to the bulk boiling, so called 'Downcomer Boiling'. Some system codes (e.g. RELAP, TRACE) predict this generated steam prevents the penetration of safety injection water into core and eventually degrades the core cooling capability. In this concern, separate effect tests on the downcomer boiling phenomena have been performed. When water in downcomer is boiled off by heated wall, interactions between void and liquid become important. Interfacial drag model is one of key factors to handle those phenomena in RELAP5/MOD3.3. So, we assessed several models related to interfacial drag in RELAP5/MOD3.3 code to obtain the most appropriate model using the experiment. EPRI and Bestion correlations are compared to Kataoka-Ishii correlation. Also, we perform the comparison by adopting Blasius model used in TRACE code. In TRACE code, Blasius model is a special interfacial drag model which applied in the downcomer only. Especially, because Bestion correlation is developed for channel having small diameter, we conduct additional assessment by multiplying factor for calibration of hydraulic diameter term within the correlation. As the results of the assessment, the modified Bestion model is most appropriate to simulate the experiments. Finally, we assess CCTF (Cylindrical Core Test Facility) C2-4 test using the improved model to confirm the validity of the developed model. (authors)

  18. Boiling liquid cauldron status report

    SciTech Connect

    Hoffman, M.A.

    1980-12-28

    The progress made over the past year in assessing the feasibility of the high-temperature, boiling cauldron blanket concept for the tanden mirror reactor is reviewed. The status of the proposed experiments and recently revised estimates of the vapor void fraction in the boiling pool are discussed.

  19. 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.

  20. 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.

  1. Pressure suppression containment system for boiling water reactor

    DOEpatents

    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.

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. Optimal control of a boiling water reactor in load-following via multilevel methods

    SciTech Connect

    Lin, C.; Grossman, L.M.

    1986-04-01

    A multilevel method is applied to the load-following control of a boiling water reactor using a nodal reactor model with practical operational constraints and thermal limits. Due to the very large size of the problem, a decomposition is made using hierarchical control techniques. The optimization of the resulting subproblems is performed using the feasible direction method. An objective functional, of quadratic form, is defined to reflect the control objective, namely to achieve the desired thermal power (tracking) with minimum effort, returning to the initial xenon and iodine concentration as closely as possible. Nodal source equation and discretized Xe-I dynamic equations are formulated as equality constraints, while the linear heat generation rate and the rate of power increase are formulated as inequality constraints. Core flow and control rod position are the control variables. A simplified model of the core is used, with 4 x 4 fuel assemblies that have one control rod at the center.

  8. 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.

  9. 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.

  10. Corrosion product deposits on boiling-water reactor cladding: Experimental and theoretical investigation of magnetic properties

    NASA Astrophysics Data System (ADS)

    Orlov, A.; Degueldre, C.; Wiese, H.; Ledergerber, G.; Valizadeh, S.

    2011-09-01

    Recent Eddy current investigations on the cladding of nuclear fuel pins have shown that the apparent oxide layers are falsified due to unexpected magnetic properties of corrosion product deposits. Analyses by Scanning Electron Microscopy (SEM) or Electron Probe Micro Analysis (EPMA) demonstrated that the deposit layer consists of complex 3-d element oxides (Ni, Mn, Fe) along with Zn, since the reactor operates with a Zn addition procedure to reduce buildup of radiation fields on the recirculation system surfaces. The oxides crystallise in ferritic spinel structures. These spinels are well-known for their magnetic behaviour. Since non-magnetic zinc ferrite (ZnFe 2O 4) may become magnetic when doped with even small amounts of Ni and/or Mn, their occurrence in the deposit layer has been analyzed. The magnetic permeability of zinc ferrite, trevorite and jacobsite and their solid solutions are estimated by magnetic moment additivity. From the void history examination, the low elevation sample (810 mm) did not face significant boiling during the irradiation cycles suggesting growth of (Mn0.092+Zn0.752+Fe0.293+)[(Fe1.713+Mn0.032+Ni0.132+)O] crystals with theoretical value of the magnetic permeability for the averaged heterogeneous CRUD layer of 9.5 ± 3. Meanwhile, (Mn0.162+Zn0.552+Fe0.293+)[(Fe1.713+Mn0.042+Ni0.252+)O] crystallizes at the mid elevation (1810 mm) with theoretical magnetic permeability for the CRUD layer of 4.2 ± 1.5 at the investigated azimuthal location. These theoretical data are compared with the magnetic permeability of the corrosion product deposited layers gained from reactor pool side Eddy current (EC) analyses (9.0 ± 1.0 for low and 3.5 ± 1.0 for high elevation). The calculated thicknesses and magnetic permeability values of the deposition layers (estimated by MAGNACROX multifrequency EC method) match together with these estimated using an "ion magnetic moment additivity" model.

  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. 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.

  13. 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.

  14. Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

    SciTech Connect

    Holcombe, S.; Eitrheim, K.; Svaerd, S. J.; Hallstadius, L.; Willman, C.

    2012-07-01

    Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

  15. 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.

  16. 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

  17. 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.

  18. Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

    NASA Astrophysics Data System (ADS)

    Gorman, Phillip Michael

    The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was

  19. 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

  20. 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.

  1. 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.

  2. On the interpretation of ECP data from operating boiling water reactors

    SciTech Connect

    Macdonald, D.D.; Urquidi-Macdonald, M.

    1995-12-31

    A method has been devised for estimating the electrochemical conditions that exist in the recirculation piping of a Boiling Water Reactor (BWR) under hydrogen water chemistry (HWC) conditions from corrosion potential measurements that are made in remote autoclaves. The technique makes use of the mixed potential model (MPM) to estimate corrosion potentials in the autoclaves and compares these estimates with measured values in an optimization on the concentrations of hydrogen peroxide and oxygen in the recirculation system. The algorithm recognizes that hydrogen peroxide decomposes in the sampling lines and that the transit times between the recirculation system and the monitoring points depend on the flow rates and sampling line diameters. An analysis is made of ECP data from three monitoring locations in the Barseback BWR in Sweden, as a function of the concentration of hydrogen in the feedwater for two flow rates (5,500 and 6,300 kg/s for the four recirculation loops).

  3. 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.

  4. A decision support system for maintenance management of a boiling-water reactor power plant

    SciTech Connect

    Shen, J.H.; Ray, A.; Levin, S.

    1996-01-01

    This article reports the concept and development of a prototype expert system to serve as a decision support tool for maintenance of boiling-water reactor (BWR) nuclear power plants. The code of the expert system makes use of the database derived from the two BWR units operated by the Pennsylvania Power and Light Company in Berwick, Pennsylvania. The operations and maintenance information from a large number of plant equipment and sub-systems that must be available for emergency conditions and in the event of an accident is stored in the database of the expert system. The ultimate goal of this decision support tool is to identify the relevant Technical Specifications and management rules for shutting down any one of the plant sub-systems or removing a component from service to support maintenance. 6 refs., 7 figs.

  5. 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; hide

    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

  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. Spectral measurements of direct and scattered gamma radiation at a boiling-water reactor site

    NASA Astrophysics Data System (ADS)

    Block, R. C.; Preiss, I. L.; Ryan, R. M.; Vargo, G. J.

    1990-12-01

    Quantitative surveys of direct and scattered gamma radiation emitted from the steam-power conversion systems of a boiling-water reactor and other on-site radiation sources were made using a directionally shielded HPGe gamma spectrometry system. The purpose of this study was to obtain data on the relative contributions and energy distributions of direct and scattered gamma radiation in the site environs. The principal radionuclide of concern in this study is 16N produced by the 16O(n,p) 16N reaction in the reactor coolant. Due to changes in facility operation resulting from the implementation of hydrogen water chemistry (HWC), the amount of 16N transported from the reactor to the main steam system under full power operation is excepted to increase by a factor of 1.2 to 5.0. This increase in the 16N source term in the nuclear steam must be considered in the design of new facilities to be constructed on site as well as the evaluation of existing facilities with repect to ALARA (As Low As Reasonably Achievable) dose limits in unrestricted areas. This study consisted of base-line measurements taken under normal BWR chemistry conditions in October, 1987 and a corresponding set taken under HWC conditions in July, 1988. Ground-level and elevated measurements, corresponding to second-story building height, were obtained. The primary conclusion of this study is that direct radiation from the steam-power conversion system is the predominant source of radiation in the site environs of this reactor and that air scattering (i.e. skyshine) does not appear to be significant.

  8. 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.

  9. 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.

  10. 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.

  11. 3-D THERMAL EVALUATIONS FOR a FUELED EXPERIMENT in the ADVANCED TEST REACTOR

    SciTech Connect

    Ambrosek, R.G.; Chang, G.S.; Utterbeck, D.J.

    2004-10-06

    The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large ''B'' experimental facility. A test configuration has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.

  12. 3-D Thermal Evaluations for a Fueled Experiment in the Advanced Test Reactor

    SciTech Connect

    Richard Ambrosek; Gray Chang; Debra Utterbeck

    2004-10-01

    The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility. A test configurations has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.

  13. Strain-induced corrosion cracking behaviour of low-alloy steels under boiling water reactor conditions

    NASA Astrophysics Data System (ADS)

    Seifert, H. P.; Ritter, S.

    2008-09-01

    The strain-induced corrosion cracking (SICC) behaviour of different low-alloy reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated boiling water reactor (BWR)/normal water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP ⩾+50 mV SHE, ⩾0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness <350 HV5) and increased (once initiated) with increasing loading rates and with increasing temperature with a possible maximum/plateau at 250 °C. A minimum KI value of 25 MPa m 1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate d KI/d t, but were not dependent on the actual KI values up to 60 MPa m 1/2. A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200-250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.

  14. 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.

  15. A two-step method for developing a control rod program for boiling water reactors

    SciTech Connect

    Taner, M.S.; Levine, S.H. ); Hsiao, M.Y. )

    1992-01-01

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift.

  16. Expert system for maintenance management of a boiling water reactor power plant

    SciTech Connect

    Hong Shen; Liou, L.W.; Levine, S.; Ray, A. ); Detamore, M. )

    1992-01-01

    An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, which must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components.

  17. Development of a MELCOR self-initialization algorithm for boiling water reactors

    SciTech Connect

    Chien, C.S.; Wang, S.J.; Cheng, S.K.

    1996-01-01

    The MELCOR code, developed by Sandia National Laboratories, is suitable for calculating source terms and simulating severe accident phenomena of nuclear power plants. Prior to simulating a severe accident transient with MELCOR, the initial steady-state conditions must be generated in advance. The current MELCOR users` manuals do not provide a self-initialization procedure; this is the reason users have to adjust the initial conditions by themselves through a trial-and-error approach. A MELCOR self-initialization algorithm for boiling water reactor plants has been developed, which eliminates the tedious trial-and-error procedures and improves the simulation accuracy. This algorithm adjusts the important plant variable such as the dome pressure, downcomer level, and core flow rate to the desired conditions automatically. It is implemented through input with control functions provided in MELCOR. The reactor power and feedwater temperature are fed as input data. The initialization work of full-power conditions of the Kuosheng nuclear power station is cited as an example. These initial conditions are generated successfully with the developed algorithm. The generated initial conditions can be stored in a restart file and used for transient analysis. The methodology in this study improves the accuracy and consistency of transient calculations. Meanwhile, the algorithm provides all MELCOR users an easy and correct method for establishing the initial conditions.

  18. Optimization of Boiling Water Reactor Loading Pattern Using Two-Stage Genetic Algorithm

    SciTech Connect

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2002-10-15

    A new two-stage optimization method based on genetic algorithms (GAs) using an if-then heuristic rule was developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). In the first stage, the LP is optimized using an improved GA operator. In the second stage, an exposure-dependent control rod pattern (CRP) is sought using GA with an if-then heuristic rule. The procedure of the improved GA is based on deterministic operators that consist of crossover, mutation, and selection. The handling of the encoding technique and constraint conditions by that GA reflects the peculiar characteristics of the BWR. In addition, strategies such as elitism and self-reproduction are effectively used in order to improve the search speed. The LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and constraints dependent on three dimensions have always necessitated the use of three-dimensional core simulators for BWRs, so that optimization of computational efficiency is required. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant in two phases. One phase is only LP optimization applying the Haling technique. The other phase is an LP optimization that considers the CRP during reactor operation. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.

  19. The Neutronics Design and Analysis of a 200-MW(electric) Simplified Boiling Water Reactor Core

    SciTech Connect

    Tinkler, Daniel R.; Downar, Thomas J.

    2003-06-15

    A 200-MW(electric) simplified boiling water reactor (SBWR) was designed and analyzed under sponsorship of the U.S. Department of Energy Nuclear Energy Research Initiative program. The compact size of a 200-MW(electric) reactor makes it attractive for countries with a less well developed engineering infrastructure, as well as for developed countries seeking to tailor generation capacity more closely to the growth of their electricity demand. The 200-MW(electric) core design reported here is based on the 600-MW(electric) General Electric SBWR core, which was first analyzed in the work performed here in order to qualify the computer codes used in the analysis. Cross sections for the 8 x 8 fuel assembly design were generated with the HELIOS lattice physics code, and core simulation was performed with the U.S. Nuclear Regulatory Commission codes RELAP5/PARCS. In order to predict the critical heat flux, the Hench-Gillis correlation was implemented in the RELAP5 code. An equilibrium cycle was designed for the 200-MW(electric) core, which provided a cycle length of more than 2 yr and satisfied the minimum critical power ratio throughout the core life.

  20. Recriticality in a BWR (boiling water reactor) following a core damage event

    SciTech Connect

    Scott, W.B.; Harrison, D.G.; Libby, R.A.; Tokarz, R.D. ); Wooton, R.D.; Denning, R.S.; Tayloe, R.W. Jr. )

    1990-12-01

    This report describes the results of a study conducted by Pacific Northwest Laboratory to assist the US Nuclear Regulatory Commission in evaluating the potential for recriticality in boiling water reactors (BWRs) during certain low probability severe accidents. Based on a conservative bounding analysis, this report concludes that there is a potential for recriticality in BWRs if core reflood occurs after control blade melting has begun but prior to significant fuel rod melting. However, a recriticality event will most likely not generate a pressure pulse significant enough to fail the vessel. Instead, a quasi-steady power level would result and the containment pressure and temperature would increase until the containment failure pressure is reached, unless actions are taken to terminate the event. Two strategies are identified that would aid in regaining control of the reactor and terminate the recriticality event before containment failure pressures are reached. The first strategy involves initiating boration injection at or before the time of core reflood if the potential for control blade melting exists. The second strategy involves initiating residual heat removal suppression pool cooling to remove the heat load generated by the recriticality event and thus extend the time available for boration. 31 figs., 17 tabs.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. Experimental studies of heat exchange for sodium boiling in the fuel assembly model: Safety substantiation of a promising fast reactor

    NASA Astrophysics Data System (ADS)

    Khafizov, R. R.; Poplavskii, V. M.; Rachkov, V. I.; Sorokin, A. P.; Trufanov, A. A.; Ashurko, Yu. M.; Volkov, A. V.; Ivanov, E. F.; Privezentsev, V. V.

    2017-01-01

    Numerical simulation of the ULOF-type accident development in a fast reactor with sodium coolant performed using the COREMELT code indicates that sodium boiling in the active core takes place. The boiling is accompanied by oscillations of the technological parameters of the reactor installation; these oscillations can go on during several tens of seconds. In this case, it is possible that a stable regime of removal of heat from residual energy release is implemented. The model of the two-phase coolant flow applied in the code has an important effect on the numerical results; that is why this model needs experimental verification. For eliminating the development of an accident resulting in destruction of the active core elements, a structural solution is proposed; the essence of it is the application of the sodium void above the reactor active core. The experimental installation was developed and the heat exchange at sodium boiling in the model fuel assembly of the fast reactor in the regimes of natural and forced circulation in the presence of the sodium void and the top end shield was studied. It was demonstrated that, in the presence of the sodium void, it is possible to provide long-term cooling of the fuel assembly for a thermal flux density on the fuel element simulator surface of up to 140 and 170 kW/m2 in the natural and forced circulation modes, respectively. The obtained data are used for more precise determination of the numerical model of sodium boiling in the fuel assembly and verification of the COREMELT code.

  9. 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.

  10. 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.

  11. 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.

  12. Pool film boiling experiments on a wire in low gravity: preliminary results.

    PubMed

    Di Marco, P; Grassi, W; Trentavizi, F

    2002-10-01

    This paper reports preliminary results for pool film boiling on a wire immersed in almost saturated FC72 recently obtained during an experimental campaign performed in low gravity on the European Space Agency Zero-G airplane, (reduced gravity level 10(-2)). This is part of a long-term research program on the effect of gravitational and electric forces on boiling. The reported data set refers to experiments performed under the following conditions: (1) Earth gravity without electric field, (2) Earth gravity with electric field, (3) low gravity without electric field, and (4) low gravity with electric field. Although a decrease of gravity causes a heat transfer degradation, the electric field markedly improves heat exchange. This improvement is so effective that, beyond a certain field value, the heat flux is no longer sensitive to gravity. Two main film boiling regimes have been identified, both in normal and in low gravity: one is affected by the electric field and the other is practically insensitive to the field influence.

  13. Prediction of dryout performance for boiling water reactor fuel assemblies based on subchannel analysis with the RINGS code

    SciTech Connect

    Knabe, P.; Wehle, F.

    1995-12-01

    A fuel assembly with a large critical power margin introduces flexibility into reload fuel management. Therefore, optimization of the bundle and spacer geometry to maximize the bundle critical power is an important design objective. With a view to reducing the extent of the complex full-scale tests usually carried out to determine the thermal-hydraulic characteristics of various assembly geometries, the subchannel analysis method was further developed with the Siemens RINGS code. The annular flow code predicts dryout power and dryout location by calculating the conditions at which the liquid film flow rate is reduced to zero, allowing for evaporation, droplet entrainment, and droplet deposition. Appropriate attention is paid to the modeling of spacer effects. Comparison with experimental data of 3 x 3 and 4 x 4 tests shows the capability of RINGS to predict the flow quality and mass flux in subchannels under typical boiling water reactor operating conditions. By using the RINGS code, experimental critical power data for 3 x 3, 4 x 4, 5 x 5, 7 x 7, 8 x 8, 9 x 9, and 10 x 10 fuel assemblies were successfully postcalculated.

  14. 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

  15. 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.

  16. 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)

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

    DOE PAGES

    Forrest, Eric C.; Don, Sarah M.; Hu, Lin -Wen; ...

    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

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

    SciTech Connect

    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 750 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.

  19. Factors influencing the precoat filtration of boiling water reactor water streams

    SciTech Connect

    Hermansson, H.P. ); Persson, G. ); Reinvall, A. )

    1994-10-01

    A series of studies on precoat filtration were carried out on condensate and preheater drains in the Swedish and Finnish boiling water reactors (BWRs). The goal was to increase knowledge about the precoat filtration process and to find physical and chemical means to improve the performance of the precoat filters in the condensate polishing plants. To achieve this goal a number of parameters, such as type of resin, bed depth, pH, oxygen and organic contaminant concentrations (measured total organic carbon), and corrosion product particle characteristics, were selected for the study. The work was mainly carried out in the power plants using an experimental facility fed with on-line sampled condensates and drains taken from the plant sampling lines. The main results are that there is a varying influence on precoat filtration from all the aforementioned parameters. The oxygen concentration, the concentration of organic contaminants, and the type of corrosion products are, however, the factors that have the strongest influence within the parameter ranges that are representative for BWR operation. The results are rather similar when the different units are compared. There are, however, some differences that could be mainly attributed to deviations in operation parameters and the subsequent differences in the corrosion product spectra. The mechanism for precoat filtration of corrosion products in BWR condensate is complex. The filtration behavior is to a large extent governed by competition between depth filtration and electrostatic interactions. During the early stages of the filtration cycle, electrostatic interaction is of great importance, whereas depth filtration becomes more important with increasing operating time. Rapid pressure drop buildup rates have been demonstrated to be caused by the presence of amorphous corrosion products. An effect from the presence of organic contaminants has been found, although this should be of little significance.

  20. Experimental Study on Thermal-Hydraulics During Start-Up in the Natural Circulation Boiling Water Reactor Concept: Effects of System Pressure and Increasing Heat Flux on the Geysering and Density Wave Oscillation

    SciTech Connect

    Hadid Subki, M.; Masanori Aritomi; Noriyuki Watanabe; Chaiwat Muncharoen

    2002-07-01

    The feasibility study in thermal-hydraulics for the future light water reactor concept is carried out. One of the essential studies is the two-phase flow instability during start-up in the natural circulation boiling water reactor (BWR) concept. It is anticipated that the occurrence of the two-phase flow instabilities during start-up significantly affects the feasibility concept, since it would cause the complexity in raising and maneuvering the power output. The purpose of the current study is to experimentally investigate the driving mechanism of the geysering and density wave oscillation in the natural circulation loop, induced by a range of system operating pressure and increasing heat flux in vertical parallel channels. The pressure range of atmospheric up to about 4 bars, and the input heat flux range of 0 up to 577 kW/m{sup 2} are applied in these experiments. An experimental apparatus of twin boiling upflow channels to simulate natural circulation flow loop has been designed, constructed and operated. The natural circulation in the loop occurs due to the density difference between two-phase region in the channels and the single-phase liquid in the downcomer. The objective of the study is to propose a rational start-up procedure in which the geysering and density wave oscillation can be prevented during startup, according to its system pressure and heat flux. Previous studies have clarified that three (3) kinds of thermo-hydraulics instabilities may occur during start-up in the natural circulation BWR depending on its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation induced by hydrostatic head fluctuation in steam separator, and (3) density wave oscillation. (authors)

  1. 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

  2. Void fraction distribution in a boiling water reactor fuel assembly and the evaluation of subchannel analysis codes

    SciTech Connect

    Inoue, Akira; Futakuchi, Masanobu; Yagi, Makoto; Mitsutake, Toru; Morooka, Shinichi

    1995-12-01

    Void fraction measurement tests for boiling water reactor (BWR) simulated nuclear fuel assemblies have been conducted using an X-ray computed tomography scanner.there are two types of fuel assemblies concerning water rods. One fuel assembly has two water rods; the other has one large water rod. The effects of the water rods on radial void fraction distributions are measured within the fuel assemblies. The results show that the water rod effect does not make a large difference in void fraction distribution. The subchannel analysis codes COBRA/BWR and THERMIT-2 were compared with subchannel-averaged void fractions. The prediction accuracy of COBRA/BWR and THERMIT-2 for the subchannel-averaged void fraction was {Delta}{alpha} = {minus}3.6%, {sigma} = 4.8% and {Delta}{alpha} = {minus}4.1%, {sigma} = 4.5%, respectively, where {Delta}{alpha} is the average of the difference measured and calculated values. The subchannel analysis codes are highly applicable for the prediction of a two-phase flow distribution within BWR fuel assemblies.

  3. Qualification of JEFF3.1.1 library for high conversion reactor calculations using the ERASME/R experiment

    SciTech Connect

    Vidal, J. F.; Noguere, G.; Peneliau, Y.; Santamarina, A.

    2012-07-01

    With its low CO{sub 2} production, Nuclear Energy appears to be an efficient solution to the global warming due to green-house effect. However, current LWR reactors are poor uranium users and, pending the development of Fast Neutron Reactors, alternative concepts of PWR with higher conversion ratio (HCPWR) are being studied again at CEA, first studies dating from the middle 80's. In these French designs, low moderation ratio has been performed by tightening the lattice pitch, achieving a conversion ratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRs are characterized by a harder neutron spectrum and the calculation uncertainties on the fundamental neutronics parameters are increased by a factor 3 regarding a standard PWR lattice, due to the major contribution of the Plutonium isotopes and of the epithermal energy range to the reaction rates. In order to reduce these uncertainties, a 3-year experimental validation program called ERASME has been performed by CEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/R experiments with the Monte Carlo code TRIPOLI4 allowed the qualification of the recommended JEFF.3.1.1 library for major neutronics parameters. K{sub eff} of the MOX under-moderated lattice is over-predicted by 440 {+-} 830 pcm (2{sigma}); the conversion ratio, indicator of the good use of uranium, is also slightly over-predicted: 2 % {+-} 4 % (2{sigma}) and the same for B4C absorber rods worth and soluble boron worth, over-predicted by 2 %, both in the 2 standard deviations range. The radial fission maps of heterogeneities (water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is 1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameter poorly calculated with an overprediction of +12.4% {+-} 1.5%. ERASME/R analysis of MOX reactivity, void effect and spectral indexes will contribute to the reevaluation of {sup 241}Am and Plutonium isotopes nuclear

  4. Detecting Dark Photons with Reactor Neutrino Experiments

    NASA Astrophysics Data System (ADS)

    Park, H. K.

    2017-08-01

    We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ <1.3 ×10-5 and ɛ <2.1 ×10-5, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

  5. 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.

  6. Acoustic Analysis for a Steam Dome and Pipings of a 1,100 MWe-Class Boiling Water Reactor

    SciTech Connect

    Yasumi, Kitajima; Masanobu, Watanabe; Keiji, Matsunaga; Tsuyoshi, Hagiwara

    2006-07-01

    For the integrity evaluation of steam dryers in up-rated nuclear power plants, we have applied acoustic analysis to a nuclear power plant steam dome and main steam pipings. We have selected a 1,100 MWe-class boiling water reactor as a subject of the analysis. We have constructed a three-dimensional finite element model, and conducted acoustic analyses. The analysis result suggested that the origin of steam pressure pulsation in high frequency range was due to vortex shedding at standpipes. (authors)

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Characterization of Phases in “Crud” from Boiling Water Reactors by Transmission Electron Microscopy

    SciTech Connect

    Dawn E. Janney; Douglas L. Porter

    2007-05-01

    This paper reports phases identified in samples of crud (activated corrosion products) from two commercial boilingwater reactors using transmission and analytical electron microscopy and selected-area electron diffraction. Franklinite (ZnFe2O4) was observed in both samples. Hematite (a-Fe2O3), crystalline silica (SiO2), a fine-grained mixture of iron oxides probably including magnetite (Fe3O4), hematite (a-Fe2O3), and goethite (a-FeOOH), and an unidentified high-Ba, high-S phase were observed in one of the samples. Willemite (Zn2SiO4), amorphous silica, and an unidentified iron– chromium phase were observed in the other. Chloride-bearing phases were found in both samples, and are assumed to represent sample contaminants. Because of the small sample volumes and numbers of particles studied and the possibility of contamination, it is not clear whether the differences between the phases observed in the two crud samples represent actual differences in the assemblages formed in the reactors.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Sodium boiling experiments in bundle geometries under fast transient and steady state conditions

    NASA Astrophysics Data System (ADS)

    Peppler, F. W.

    Flow loss experiments were carried out in tubular and 7 pin test-sections and in a sodium cooled 37 pin bundle with sinusoidal axial heat flux profile. Thermohydraulic behavior under fast loss of flow conditions, particularly three-dimensional phenomena such as the development of the temperature distribution, the boiling region, the onset and extension of dryout, and the flow regimes related to different boiling periods were studied. The experimental results provide data for the validation of three-dimensional two-phase hydrodynamic codes. Stationary boiling tests provide information on two-phase pressure drop in bundles, the flow regimes and the conditions prevailing during the approach to critical heat flux conditions.

  18. 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.

  19. Benchmark evaluation of the RELAP code to calculate boiling in narrow channels

    SciTech Connect

    Kunze, J.F.; Loyalka, S.K. ); McKibben, J.C.; Hultsch, R.; Oladiran, O.

    1990-06-01

    The RELAP code has been tested with benchmark experiments (such as the loss-of-fluid test experiments at the Idaho National Engineering Laboratory) at high pressures and temperatures characteristic of those encountered in loss-of-coolant accidents (LOCAs) in commercial light water power reactors. Application of RELAP to the LOCA analysis of a low pressure (< 7 atm) and low temperature (< 100{degree}C), plate-type research reactor, such as the University of Missouri Research Reactor (MURR), the high-flux breeder reactor, high-flux isotope reactor, and Advanced Test Reactor, requires resolution of questions involving overextrapolation to very low pressures and low temperatures, and calculations of the pulsed boiling/reflood conditions in the narrow rectangular cross-section channels (typically 2 mm thick) of the plate fuel elements. The practical concern of this problem is that plate fuel temperatures predicted by RELAP5 (MOD2, version 3) during the pulsed boiling period can reach high enough temperatures to cause plate (clad) weakening, though not melting. Since an experimental benchmark of RELAP under such LOCA conditions is not available and since such conditions present substantial challenges to the code, it is important to verify the code predictions. The comparison of the pulsed boiling experiments with the RELAP calculations involves both visual observations of void fraction versus time and measurements of temperatures near the fuel plate surface.

  20. 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.

  1. Analytic solution to verify code predictions of two-phase flow in a boiling water reactor core channel

    SciTech Connect

    Chen, K.F.; Olson, C.A.

    1983-09-01

    One reliable method that can be used to verify the solution scheme of a computer code is to compare the code prediction to a simplified problem for which an analytic solution can be derived. An analytic solution for the axial pressure drop as a function of the flow was obtained for the simplified problem of homogeneous equilibrium two-phase flow in a vertical, heated channel with a cosine axial heat flux shape. This analytic solution was then used to verify the predictions of the CONDOR computer code, which is used to evaluate the thermal-hydraulic performance of boiling water reactors. The results show excellent agreement between the analytic solution and CONDOR prediction.

  2. Analytic solution to verify code predictions of two-phase flow in a boiling water reactor core channel. [CONDOR code

    SciTech Connect

    Chen, K.F.; Olson, C.A.

    1983-09-01

    One reliable method that can be used to verify the solution scheme of a computer code is to compare the code prediction to a simplified problem for which an analytic solution can be derived. An analytic solution for the axial pressure drop as a function of the flow was obtained for the simplified problem of homogeneous equilibrium two-phase flow in a vertical, heated channel with a cosine axial heat flux shape. This analytic solution was then used to verify the predictions of the CONDOR computer code, which is used to evaluate the thermal-hydraulic performance of boiling water reactors. The results show excellent agreement between the analytic solution and CONDOR prediction.

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. Neutrino oscillation experiments at nuclear reactors

    NASA Astrophysics Data System (ADS)

    Grassi, Marco

    2000-08-01

    The current status of the search for neutrino oscillations at reactors is reviewed, with a particular emphasis given to the final results recently published by the CHOOZ experiment. The results of the Bugey experiments and the status of the Palo Verde experiment are also discussed.

  8. 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.

  9. Cold neutron tomography of annular coolant flow in a double subchannel model of a boiling water reactor

    NASA Astrophysics Data System (ADS)

    Kickhofel, J. L.; Zboray, R.; Damsohn, M.; Kaestner, A.; Lehmann, E. H.; Prasser, H.-M.

    2011-09-01

    Dryout of the liquid coolant film on fuel pins at the top of boiling water reactor (BWR) cores constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is a limiting factor in the thermal power, and therefore the economy, of BWRs. Ongoing research on multiphase annular flow, specifically the liquid film thickness, is fundamental not only to nuclear reactor safety and operation but also to that of evaporators, condensers, and pipelines in a general industrial context. We have performed cold neutron tomography of adiabatic air water annular flow in a scaled up model of the subchannel geometry found in BWR fuel assemblies today. All imaging has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institut in Switzerland. Neutron tomography is shown to excel in investigating the interactions of air water two phase flows with spacer vanes of different geometry. The high resolution, high contrast measurements provide spatial distributions of the coolant on top of the surfaces of the spacer, including the vanes, and in the subchannel downstream of the spacers.

  10. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    SciTech Connect

    Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  11. 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.

  12. 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)

  13. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    SciTech Connect

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  14. 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.

  15. A proof-of-concept transient diagnostic expert system for BWRs (Boiling Water Reactors)

    SciTech Connect

    Yoshida, K.; Naser, J.A.

    1988-05-01

    A proof-of-concept transient diagnostic expert system has been developed to identify the cause and the type of an abnormal transient in a boiling water nuclear power plant. For this expert system development, the calculational results of the simulation code RETRAN were used as the knowledge source. The knowledge extracted from the RETRAN analyses was transformed into IF-THEN rules in the knowledge base for the expert system. An important feature of this expert system is the introduction of certainty factors to allow diagnosis even in the cases where data may be either missing or marked as invalid. To increase the capability of this diagnostic system to distinguish between similiar transients, backward chaining reasoning is used to support the forward chaining reasoning with certainty factors. Through this effort, it has been demonstrated that an expert system can be successfully used to create a transient diagnostic system. It has also successfully demonstrated that RETRAN can be used as the knowledge source for developing the knowledge base of the diagnostic system.

  16. The first results of He II boiling visualization experiment conducted under 4.7 seconds microgravity conditions

    NASA Astrophysics Data System (ADS)

    Takada, S.; Kimura, N.; Pietrowicz, S.; Grunt, K.

    2017-09-01

    An experiment regarding boiling of superfluid helium (He II) has been carried out under conditions of microgravity, in order to investigate the dynamics of the phase transition. A small cryostat equipped with visualization setup has been utilized for this purpose. Presence of two orthogonal optical axes allowed for registering of 3-dimensional images of a vapor bubble induced by a micro heater. Microgravity environment has been produced for about 4.7 s using the 122 m high drop tower facility at ZARM (Center of Applied Space Technology and Microgravity), University of Bremen, Germany. The experimental campaign consisting of 32 drops has been successfully conducted, while avoiding any damage to the equipment.

  17. PROSPECT: The Precision Reactor Oscillation and Spectrum Experiment

    NASA Astrophysics Data System (ADS)

    Langford, Thomas; Prospect Collaboration

    2016-09-01

    PROSPECT is a phased experiment consisting of segmented 6Li-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 7-12 m from the compact, highly enriched uranium core. Over the past three years, PROSPECT has deployed multiple 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. PROSPECT will probe the sterile neutrino best-fit region at 3 σ within one year of operation at HFIR. We will discuss the design, experimental program, and discovery potential of the experiment.

  18. Fast critical experiment data for space reactors

    SciTech Connect

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

    1987-01-01

    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, the old experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, we have calculated about 20 fast benchmark critical experiments 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.

  19. 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.

  20. 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.

  1. Experimental study of the effect of void reactivity feedback on the behavior of the scaled model boiling water reactor

    NASA Astrophysics Data System (ADS)

    Meftah, Khaled

    A Scaled Model Boiling Water Reactor (SMBWR) model uses low pressure (i.e., 0.095 MPa) water in a heated channel 0.5 meters in length with four electrically heated fuel simulator rods. The axial void profile in the channel is measured using conductivity probes and the power to the heaters is modulated according to the void fraction to simulate void reactivity feedback. The steam from the heated channel is passed through a valve that reduces the pressure to 0.012 MPa where the steam is condensed in conditions similar to those found in a conventional BWR condenser. The feedwater flow rate, heater power, and instrumentation in the facility are controlled and monitored through a Quadra 950 computer running LabVIEW software. The void fraction signals are analyzed to identify the different flow regimes and determine the vapor velocity in the SMBWR channel using features of the probability density function and power spectral density. The void coefficient of reactivity is modified in the BWR scale model through the LabVIEW interface and the effect on the behavior of the channel is directly observed. The system response is reported for abrupt stepwise pressure changes and abrupt stepwise power changes. The response is typical of that expected for a BWR. The void reactivity feedback effect is also examined by analyzing the frequency response of the channel void fraction at steady state.

  2. 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.

  3. 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).

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Effect of boiling on the content of ascorbigen, indole-3-carbinol, indole-3-acetonitrile, and 3,3'-diindolylmethane in fermented cabbage.

    PubMed

    Ciska, Ewa; Verkerk, Ruud; Honke, Joanna

    2009-03-25

    The aim of the study was to investigate the effect of the boiling process on the content of ascorbigen, indole-3-carbinol, indole-3-acetonitrile, and 3,3'-diindolylmethane in fermented cabbage. The cabbage was boiled for 5 to 60 min. Boiling resulted in a decrease of the total content of the compounds analysed. The changes were mainly caused by leaching of ascorbigen predominating in cabbage into cooking water and by its thermal hydrolysis. Ascorbigen losses resulting from thermal hydrolysis accounted for 30% after 10 min of boiling and for 90% after 60 min of boiling. One of the ascorbigen breakdown products was indole 3 carbinol; the decrease in ascorbigen content was accompanied by a drastic increase in the content of 3,3'-diindolylmethane, a condensation product of indole-3-carbinol. After 40 and 50 min of boiling, the total content of 3,3'-diindolylmethane in cabbage and cooking water was approximately 0.2 micromol/100 g and was 6-fold higher than that in uncooked cabbage. 3,3'-Diindolylmethane synthesis proceeded within the plant tissue. After 10 min of boiling, the content of free indole-3-carbinol and indole-3-acetonitrile stabilized at the level of about 80% as compared to the uncooked cabbage.

  15. Dynamics of liquid helium boil-off experiments with a step change in pressure

    NASA Astrophysics Data System (ADS)

    Cha, Y. S.; Niemann, R. C.; Hull, J. R.

    The results of dynamic analysis of the effect of pressure variations during helium boil-off experiments are presented. A general solution of the diffusion equation with a time-dependent boundary condition is employed to describe the dynamic response of the liquid helium system under variable pressure conditions, and a solution is obtained for the special case when the system is subjected to a step change in pressure. The calculated temperature response of the liquid indicates that most of the experiments were not likely to have reached equilibrium as a result of the low thermal diffusivity of liquid helium. The initial rate of evaporation or condensation is large, and the rate decreases sharply with time. A method is proposed to account for the transient effect that is observed during calculation of the heat loss rate from a helium boil-off experiment. By assuming that there is no mixing at all, the present analysis provides an estimate of the upper (condensation) or lower (evaporation) bound of the heat loss rate as a result of a pressure increase or decrease in the system. A previously reported equilibrium analysis is expected to apply to situations where complete mixing occurred in the bulk liquid and provides the opposite limits.

  16. Performance Evaluation of the International Space Station Flow Boiling and Condensation Experiment (FBCE) Test Facility

    NASA Technical Reports Server (NTRS)

    Hasan, Mohammad; Balasubramaniam, R.; Nahra, Henry; Mackey, Jeff; Hall, Nancy; Frankenfield, Bruce; Harpster, George; May, Rochelle; Mudawar, Issam; Kharangate, Chirag R.; hide

    2016-01-01

    A ground-based experimental facility to perform flow boiling and condensation experiments is built in support of the development of the long duration Flow Boiling and Condensation Experiment (FBCE) destined for operation on board of the International Space Station (ISS) Fluid Integrated Rack (FIR). We performed tests with the condensation test module oriented horizontally and vertically. Using FC-72 as the test fluid and water as the cooling fluid, we evaluated the operational characteristics of the condensation module and generated ground based data encompassing the range of parameters of interest to the condensation experiment to be performed on the ISS. During this testing, we also evaluated the pressure drop profile across different components of the fluid subsystem, heater performance, on-orbit degassing subsystem, and the heat loss from different components. In this presentation, we discuss representative results of performance testing of the FBCE flow loop. These results will be used in the refinement of the flight system design and build-up of the FBCE which is scheduled for flight in 2019.

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

    PubMed Central

    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

  18. 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.

  19. Pool Boiling with Non-condensable Gas in Microgravity: Results of a Sounding Rocket Experiment

    NASA Astrophysics Data System (ADS)

    Kannengieser, Olivier; Colin, Catherine; Bergez, Wladimir

    2010-09-01

    Pool boiling experiments in microgravity have been performed in the Sounding Rocket Maser 11. A heated plate of 1 cm 2 was located at the bottom of a small cylindrical tank partly filled with a refrigerant Novec HFE7000 pressurized with Nitrogen. Experiments were performed at different reservoir pressures and wall heat fluxes. The wall heat flux and wall temperature were simultaneously measured during the experiment and the behavior of the bubbles on the heater was filmed with a video camera through the transparent wall of the reservoir. The presence of Nitrogen dissolved inside the liquid led to a strong Marangoni convection around the bubble. The effect of Marangoni convection and evaporation on the wall heat transfer is analyzed in function of the relative values of the wall temperature and saturation temperature.

  20. 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.

  1. 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)

  2. Using reactor operating experience to improve the design of a new Broad Application Test Reactor

    SciTech Connect

    Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

    1993-07-01

    Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

  3. 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.

  4. 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.

  5. Subchannel analysis applied to neutron noise investigation in a boiling water reactor

    SciTech Connect

    Khan, H.J.

    1986-01-01

    Drift flux parameters were developed for subchannel configuration of BWR rod bundle geometry in order to calculate the void fraction and void propagation velocities inside the individual subchannels. Void fraction calculations were performed using subchannel drift flux code CANAL. Using void and flow distributions, values of the distribution parameters were estimated for bundle averaged void fraction calculation. The simple geometries taken into consideration are circular tubes, rectangular channels, and parallel plates. A C/sub 0/ model was also developed for parallel subchannel configuration. Successful prediction of average void fraction is observed for the annular rod bundle geometry of the FRIGG experiment. A C/sub 0/ model developed for one-dimensional subchannel geometry was applied to annular subchannels using Zuber-Findlay's void quality model. This is the first time subchannel void fraction is predicted by subchannel drift flux parameters. It appears that the magnitude of C/sub 0/ varies between subchannels but remains almost constant within each subchannel. A two-dimensional derivation of C/sub 0/ for the three types of subchannel (corner, side, and center) of BWR rod-bundle geometry was developed.

  6. 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.

  7. 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…

  8. 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…

  9. 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

  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. 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.

  12. 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.

  13. The Daya Bay Reactor Electron Anti-neutrino Oscillation Experiment

    NASA Astrophysics Data System (ADS)

    Liu, Jianglai

    2007-10-01

    The phenomenon of neutrino flavor oscillations is now well-established. Mixing among the three flavors is characterized by three mixing angles, with θ13 being the only presently unknown angle. A precise measurement of θ13 using nuclear reactors as a source of electron anti-neutrinos requires high electron anti-neutrino flux, ˜2 km baselines, as well as good shielding to reduce cosmogenic backgrounds. The Daya Bay nuclear reactor complex located in south China is an ideal site to perform such a measurement. We have proposed an experiment at Daya Bay utilizing multiple baselines (between 0.3 and 2 km) and multiple liquid scintillator detector modules. Since the formal physics proposal in 2006, much progress has been made by the collaboration in the design of the experiment. The civil construction of the experiment will begin this year. In this talk, I will give an overview of the experiment, and report on the recent progress and the project status.

  14. 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.

  15. 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.

  16. Letter Report: Progress in developing EQ3/6 for modeling boiling processes

    SciTech Connect

    Wolery, T. J., LLNL

    1995-08-28

    EQ3/6 is a software package for geochemical modeling of aqueous systems, such as water/rock or waste/water rock. It is being developed for a variety of applications in geochemical studies for the Yucca Mountain Site Characterization Project. The present focus is on development of capabilities to be used in studies of geochemical processes which will take place in the near-field environment and the altered zone of the potential repository. We have completed the first year of a planned two-year effort to develop capabilities for modeling boiling processes. These capabilities will interface with other existing and future modeling capabilities to provide a means of integrating the effects of various kinds of geochemical processes in complex systems. This year, the software has been modified to allow the formation of a generalized gas phase in a closed system for which the temperature and pressure are known (but not necessarily constant). The gas phase forms when its formation is thermodynamically favored; that is, when the system pressure is equal to the sum of the partial pressures of the gas species as computed from their equilibrium fugacities. It disappears when this sum falls below that pressure. `Boiling` is the special case in which the gas phase which forms consists mostly of water vapor. The reverse process is then `condensation.` To support calculations of boiling and condensation, we have added a capability to calculate the fugacity coefficients of gas species in the system H{sub 2}O-CO{sub 2}-CH{sub 4}-H{sub 2},-Awe{sub 2}-N{sub 2},-H{sub 2}S-NH3. This capability at present is accurate only at relatively low pressures, but is adequate for all likely repository boiling conditions. We have also modified the software to calculate changes in enthalpy (heat) and volume functions. Next year we will be extending the boiling capability to calculate the pressure or the temperature at known enthalpy. We will also add an option for open system boiling.

  17. 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.

  18. Equalization of energy density in boiling water reactors (as exemplified by WB-50). Development and testing of WB -50 computational model on the basis of MCU-RR code

    NASA Astrophysics Data System (ADS)

    Chertkov, Yu B.; Disyuk, V. V.; Pimenov, E. Yu; Aksenova, N. V.

    2017-01-01

    Within the framework of research in possibility and prospects of power density equalization in boiling water reactors (as exemplified by WB-50) a work was undertaken to improve prior computational model of the WB-50 reactor implemented in MCU-RR software. Analysis of prior works showed that critical state calculations have deviation of calculated reactivity exceeding ±0.3 % (ΔKef/Kef) for minimum concentrations of boric acid in the reactor water and reaching 2 % for maximum concentration values. Axial coefficient of nonuniform burnup distribution reaches high values in the WB-50 reactor. Thus, the computational model needed refinement to take into account burnup inhomogeneity along the fuel assembly height. At this stage, computational results with mean square deviation of less than 0.7 % (ΔKef/Kef) and dispersion of design values of ±1 % (ΔK/K) shall be deemed acceptable. Further lowering of these parameters apparently requires root cause analysis of such large values and paying more attention to experimental measurement techniques.

  19. The Zero Boil-Off Tank Experiment Ground Testing and Verification of Fluid and Thermal Performance

    NASA Technical Reports Server (NTRS)

    Chato, David J.; Kassemi, Mohammad; Kahwaji, Michel; Kieckhafer, Alexander

    2016-01-01

    The Zero Boil-Off Technology (ZBOT) Experiment involves performing a small scale International Space Station (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 work that has been done to demonstrate that the ZBOT experiment is capable of performing the functions required to produce a meaningful and accurate results, prior to its launch to the International Space Station. Main systems discussed are expected to include the thermal control system, the optical imaging system, and the tank filling system.This work is sponsored by NASAs Human Exploration Mission Directorates Physical Sciences Research program.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. Microgravity experiments on boiling and applications: research activity of advanced high heat flux cooling technology for electronic devices in Japan.

    PubMed

    Suzuki, Koichi; Kawamura, Hiroshi

    2004-11-01

    Research and development on advanced high heat flux cooling technology for electronic devices has been carried out as the Project of Fundamental Technology Development for Energy Conservation, promoted by the New Energy and Industrial Technology Development Organization of Japan (NEDO). Based on the microgravity experiments on boiling heat transfer, the following useful results have obtained for the cooling of electronic devices. In subcooled flow boiling in a small channel, heat flux increases considerably more than the ordinary critical heat flux with microbubble emission in transition boiling, and dry out of the heating surface is disturbed. Successful enhancement of heat transfer is achieved by a capillary effect from grooved surface dual subchannels on the liquid supply. The critical heat flux increases 30-40 percent more than for ordinary subchannels. A self-wetting mechanism has been proposed, following investigation of bubble behavior in pool boiling of binary mixtures under microgravity. Ideas and a new concept have been proposed for the design of future cooling system in power electronics.

  5. Fission-reactor experiments for fusion-materials research

    SciTech Connect

    Grossbeck, M.L.; Bloom, E.E.; Woods, J.W.; Vitek, J.M.; Thomas, K.R.

    1982-01-01

    The US Fusion Materials Program makes extensive use of fission reactors to study the effects of simulated fusion environments on materials and to develop improved alloys for fusion reactor service. The fast reactor, EBR-II, and the mixed spectrum reactors, HFIR and ORR, are all used in the fusion program. The HFIR and ORR produce helium from transmutations of nickel in a two-step thermal neutron absorption reaction beginning with /sup 58/Ni, and the fast neutrons in these reactors produce atomic displacements. The simultaneous effects of these phenomena produce damage similar to the very high energy neutrons of a fusion reactor. This paper describes irradiation capsules for mechanical property specimens used in the HFIR and the ORR. A neutron spectral tailoring experiment to achieve the fusion reactor He:dpa ratio will be discussed.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment

    NASA Astrophysics Data System (ADS)

    Seifert, H. P.; Ritter, S.; Shoji, T.; Peng, Q. J.; Takeda, Y.; Lu, Z. P.

    2008-08-01

    The stress corrosion cracking (SCC) and corrosion fatigue behaviour perpendicular and parallel to the fusion line in the transition region between the Alloy 182 Nickel-base weld metal and the adjacent SA 508 Cl.2 low-alloy reactor pressure vessel (RPV) steel of a simulated dissimilar metal weld joint was investigated under boiling water reactor normal water chemistry conditions. A special emphasis was placed to the question whether a fast growing interdendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic SCC crack growth was observed in high-purity or sulphate-containing oxygenated water under constant or periodical partial unloading conditions for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the interdendritic SCC crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high rate as a transgranular crack into the heat-affected zone and base metal of the adjacent low-alloy steel. The observed SCC cracking behaviour at the interface correlates excellently with the field experience of such dissimilar metal weld joints, where SCC cracking was usually confined to the Alloy 182 weld metal.

  11. 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

  12. 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)

  13. Bubble Departure Diameter and Bubble Release Frequency Measurement from TAMU Subcooled Flow Boiling Experiment

    SciTech Connect

    Yoo, Jun Soo

    2016-12-01

    The bubble departure diameter and bubble release frequency were obtained through the analysis of TAMU subcooled flow boiling experimental data. The numerous images of bubbles at departure were analyzed for each experimental condition to achieve the reliable statistics of the measured bubble parameters. The results are provided in this report with simple discussion.

  14. Pool boiling

    SciTech Connect

    Lallemand, M.

    1993-10-01

    Heat transfer between a wall and a stagnant boiling liquid is reviewed in this paper. The effect of different parameters on the boiling curve is pointed out on the basis of experimental data from the literature. Augmentation of heat transfer by enhanced surfaces is described briefly. The available correlations for prediction of heat transfer coefficients are given for the entire boiling curve, i.e., nucleate, transitional, and film boiling, and critical points. These correlations are useful for the design and operation of various heat-exchange systems.

  15. Critical heat flux and boiling heat transfer to water in a 3-mm-diameter horizontal tube.

    SciTech Connect

    Yu, W.; Wambsganss, M. W.; Hull, J. R.; France, D. M.

    2000-12-04

    Boiling of the coolant in an engine, by design or by circumstance, is limited by the critical heat flux phenomenon. As a first step in providing relevant engine design information, this study experimentally addressed both rate of boiling heat transfer and conditions at the critical point of water in a horizontal tube of 2.98 mm inside diameter and 0.9144 m heated length. Experiments were performed at system pressure of 203 kPa, mass fluxes in range of 50 to 200 kg/m{sup z}s, and inlet temperatures in range of ambient to 80 C. Experimental results and comparisons with predictive correlations are presented.

  16. 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.

  17. Film boiling on spheres in single- and two-phase flows. Final report

    SciTech Connect

    Liu, C.; Theofanous, T.G.

    1994-12-01

    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{degrees}C), liquid velocity (from 0 to 2 m/s), sphere superheat (from 200 to 900{degrees}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 {alpha} 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 multisphere structure on the film boiling heat transfer in single- and two-phase flows.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Recent results of Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    Leitner, R.; Daya Bay Collaboration

    2017-04-01

    The Daya Bay reactor neutrino experiment has been designed to precisely measure the least known neutrino mixing angle θ13. In March 2012, Daya Bay collaboration announced [Daya Bay Collaboration (F. P. An et al.), Observation of electron-antineutrino disappearance at Daya Bay, Phys. Rev. Lett. 108 (2012) 171803] the observation of non-zero value of sin2 ⁡ 2θ13. Because of large statistics of detected antineutrinos and excellent performance of the experiment, Daya Bay continuously improves the precision of world best measurement of sin2 ⁡ 2θ13. In addition it provides results on neutrino mass splitting Δmee2 competitive with measurements of other experiments, results on precise measurement of reactor fluxes and on limits of the existence of hypothetical fourth neutrino. In this paper, we report the results available by the time of the 6th Capri workshop: the measurement of oscillation parameters sin2 ⁡ (2θ13) = 0.084 ± 0.005 and | Δ mee2 | = (2.42 ± 0.11) ×10-3eV2 [Daya Bay Collaboration (F. P. An et al.), New Measurement of Antineutrino Oscillation with the Full Detector Configuration at Daya Bay, Phys. Rev. Lett. 115 (2015) no. 11, 111802], searches for sterile neutrinos [Daya Bay Collaboration (F. P. An et al.) Search for a Light Sterile Neutrino at Daya Bay, Phys. Rev. Lett. 113 (2014) 141802] and precise measurement of reactor neutrino flux [Daya Bay Collaboration (F. P. An et al.), Measurement of the Reactor Anti-neutrino Flux and Spectrum at Daya Bay, Phys. Rev. Lett. 116 (2016) no. 6, 061801]. These are based on 621 days of measurement, most of the data has been taken in full detector configuration. More precise results [Daya Bay Collaboration (F. P. An et al.), Measurement of electron antineutrino oscillation based on 1230 days of operation of the Daya Bay experiment, arxiv:arXiv:1610.04802] with 1230 days of operation have been presented few weeks later at the Neutrino 2016 conference.

  3. 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.

  4. 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.

  5. 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.

  6. Double Chooz and a history of reactor θ13 experiments

    DOE PAGES

    Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

    2016-04-11

    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. In addition, the DC group presented an indication of disappearance of the reactor neutrinos at a baseline of similar to 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 naturemore » 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.« less

  7. The application of the high-speed photography in the experiments of boiling liquid expanding vapor explosions

    NASA Astrophysics Data System (ADS)

    Chen, Sining; Sun, Jinhua; Chen, Dongliang

    2007-01-01

    The liquefied-petroleum gas tank in some failure situations may release its contents, and then a series of hazards with different degrees of severity may occur. The most dangerous accident is the boiling liquid expanding vapor explosion (BLEVE). In this paper, a small-scale experiment was established to experimentally investigate the possible processes that could lead to a BLEVE. As there is some danger in using LPG in the experiments, water was used as the test fluid. The change of pressure and temperature was measured during the experiment. The ejection of the vapor and the sequent two-phase flow were recorded by a high-speed video camera. It was observed that two pressure peaks result after the pressure is released. The vapor was first ejected at a high speed; there was a sudden pressure drop which made the liquid superheated. The superheated liquid then boiled violently causing the liquid contents to swell, and also, the vapor pressure in the tank increased rapidly. The second pressure peak was possibly due to the swell of this two-phase flow which was likely to violently impact the wall of the tank with high speed. The whole evolution of the two-phase flow was recorded through photos captured by the high-speed video camera, and the "two step" BLEVE process was confirmed.

  8. The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

    SciTech Connect

    Hagen, T.H.J.J. van der; Stekelenburg, A.J.C.

    1995-09-01

    The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

  9. E-chem page: A Support System for Remote Diagnosis of Water Quality in Boiling Water Reactors

    SciTech Connect

    Naohiro Kusumi; Takayasu Kasahara; Kazuhiko Akamine; Kenji Tada; Naoshi Usui; Nobuyuki Oota

    2002-07-01

    It is important to control and maintain water quality for nuclear power plants. Chemical engineers sample and monitor reactor water from various subsystems and analyze the chemical quality as routine operations. With regard to controlling water quality, new technologies have been developed and introduced to improve the water quality from both operation and material viewpoints. To maintain the quality, it is important to support chemical engineers in evaluating the water quality and realizing effective retrieval of stored data and documents. We have developed a remote support system using the Internet to diagnose BWR water quality, which we call e-chem page. The e-chem page integrates distributed data and information in a Web server, and makes it easy to evaluate the data on BWR water chemistry. This system is composed of four functions: data transmission, water quality evaluation, inquiry and history retrieval system, and reference to documents on BWR water chemistry. The developed system is now being evaluated in trial operations by Hitachi, Ltd. and an electric power company. In addition diagnosis technology applying independent component analysis (ICA) is being developed to improve predictive capability of the system. This paper describes the structure and function of the e-chem page and presents results of obtained with the proposed system for the prediction of chemistry conditions in reactor water. (authors)

  10. Initial Experiments on Fuzzy Control for Nuclear Reactor Operations at the Belgian Reactor 1

    SciTech Connect

    Da Ruan

    2003-08-15

    The application of fuzzy logic control (FLC) in the domain of the nuclear industry presents a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear reactors and the very strict safety regulations in force for nuclear power plants. The very same regulations prevent a researcher from quickly introducing novel control methods into this field. On the other hand, the application of FLC has, despite the ominous sound of the word 'fuzzy' to nuclear engineers, a number of very desirable advantages over classical control, e.g., its robustness and the capability to include human experience into the controller. In this paper an FLC for controlling the power level of a nuclear reactor is described. The study is intended to assess the applicability of FLC in this domain. The final goal is to develop an optimized and intrinsically safe controller. After reviewing some available literature on FLC in nuclear reactors, an FLC is proposed and first tested by comparing it with the classical controller of the Belgian reactor 1 (BR1). In the next step the BR1 at the Belgian Nuclear Research Center (SCK-CEN) was used as a test bed to implement a programmable logic controller-based hardware controller. The BR1 reactor is internationally regarded as a nuclear calibration reference. It therefore provides an excellent environment for this type of experiment because over the years considerable knowledge of the static and dynamic properties of the reactor has been accumulated. The project (1995-1999) aimed at investigating the added value and technical limits of FLC for nuclear reactor operations. The progress made in these experiments including closed-loop experiments are presented and discussed in this paper.

  11. 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.

  12. Recent Results from the Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Huang, En-Chuan

    2016-11-01

    The Daya Bay Reactor Neutrino Experiment is designed to precisely measure the mixing parameter sin2 2θ13 via relative measurements with eight functionally identical antineutrino detectors (ADs). In 2012, Daya Bay has first measured a non-zero sin2 2θ13 value with a significance larger than 5σ with the first six ADs. With the installation of two new ADs to complete the full configuration, Daya Bay has continued to increase statistics and lower systematic uncertainties for better precision of sin2 2θ13 and for the exploration of other physics topics. In this proceeding, the latest analysis results of sin2 2θ13 and |Δm 2 ee|, including a measurement made with neutron capture on Gadolinium and an independent measurement made with neutron capture on hydrogen are presented. The latest results of the search for sterile neutrino in the mass splitting range of 10-3 eV2 < |Δm 2 41| < 0.3 eV2 and the absolute measurement of the rate and energy spectrum of reactor antineutrinos will also be presented.

  13. Status of the JUNO reactor anti-neutrino experiment

    NASA Astrophysics Data System (ADS)

    Lu, Haoqi; JUNO Collaboration

    2017-06-01

    The Jiangmen Underground Neutrino Observatory (JUNO) is a reactor antineutrino experiment with the aim to determine the neutrino mass hierarchy. The detector will be filled with 20 kilotons of liquid scintillator and instrumented with 18000 20-inch PMTs to achieve an unprecedented energy resolution of 3%@1 MeV. A 35.4 m diameter acrylic sphere will be built as a liquid scintillator vessel.The detector will be constructed in a 700-m-deep-underground laboratory to reduce cosmogenic muon flux. An external veto cosisting of a water Cherenkov detector and a top tracker will be used for cosmogenic muon detection and background reduction. The mass hierarchy sensitivity is expected to reach 3-4σ after 6 years of data taking. Civil construction and detector R&D are underway. Data taking is expected to start in 2020.

  14. New reactor neutrino experiments besides double-CHOOZ.

    SciTech Connect

    Goodman, M. C.; High Energy Physics

    2005-01-01

    Several new reactor neutrino experiments are being considered to measure the parameter {theta}{sub 13}. The current plans for Angra, Braidwood, Daya Bay, KASKA and KR2DET are reviewed. A case is made that, together with Double-CHOOZ, a future world program should include at least three such experiments.

  15. Background studies for the MINER Coherent Neutrino Scattering reactor experiment

    NASA Astrophysics Data System (ADS)

    Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

    2017-05-01

    The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

  16. Pulsed boiling in two narrow channels - comparison of experiment with RELAP

    SciTech Connect

    Chao, W.W.; Kunze, J.F.; Dai, W.; Loyalka, S.K. )

    1992-01-01

    Thermal-hydraulic transient/safety codes, such as RELAP, have been benchmarked for typical research reactor conditions, but such benchmarking for research reactors has not been adequate. Our simulation of the center fuel plate of the research reactor at the University of Missouri (MURR) using RELAP5/MOD2 after a hypothetical loss-of-coolant accident (LOCA) indicates that such benchmarking would be of considerable interest and necessity, as an increase in power from the current 10 MW to [approx]30 MW is planned. There is no safety concern at a steady-state power of 30 MW. However, the application of RELAP at low pressure (1 atm), low temperature (< 400K), and for plate-type flow during a LOCA with flow reversal occurs from the normal downflow to upward convectional flow, which indicates a region of concern.

  17. Experiment on Flow Boiling of HFC134a in a Multi-port Extruded Tube

    NASA Astrophysics Data System (ADS)

    Kuwahara, Ken; Koyama, Shigeru; Kazari, Kengo; Nakahita, Kouichi

    In the present study, the local heat transfer and pressure drop characteristics are investigated experimentally for the flow boiling of refrigerant HFC134a in a multi-port extruded tube of 1.06mm in hydraulic diameter. The test tube is 865mm in total length made of aluminum. The pressure drop was measured at an interval of 191mm, the local heat transfer coefficient was measured in every subsection of 75mm in effective heating length. The measured pressure drop was compared with a few correlations. Those data agree with the correlation of Koyama et al., proposed for condensation process, and of Friedel. The local heat transfer coefficients were also compared with correlation of Yu et al. It is pointed out that multiplier of the convection term in a rectangular channel is different from that of circular tube.

  18. 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.

  19. 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.

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. Investigating the spectral anomaly with different reactor antineutrino experiments

    NASA Astrophysics Data System (ADS)

    Buck, C.; Collin, A. P.; Haser, J.; Lindner, M.

    2017-02-01

    The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in 235U with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted 235U spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

  6. Core Plasma Characteristics of a Spherical Tokamak D-3He Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Shi, Bingren

    2005-04-01

    The magnetic fusion reactor using the advanced D-3He fuels has the advantage of much less-neutron productions so that the consequent damages to the first wall are less serious. If the establishment of this kind of reactor becomes realistic, the exploration of 3He on the moon will be largely motivated. Based on recent progresses in the spherical torus (ST) research, we have physically designed a D-3He fusion reactor using the extrapolated results from the ST experiments and also the present-day tokamak scaling. It is found that the reactor size significantly depends on the wall reflection coefficient of the synchrotron radiation and of the impurity contaminations. The secondary reaction between D-D that promptly leads to the D-T reaction producing 14 MeV neutrons is also estimated. Comparison of this D-3He ST reactor with the D-T reactor is made.

  7. 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.

  8. ORNL rod-bundle heat-transfer test data. Volume 2. Thermal-Hydraulic Test Facility experimental data report for test 3. 03. 6AR - transient film boiling in upflow

    SciTech Connect

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

    1982-04-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) Test 3.03.6AR. This test was conducted by members of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on May 21, 1980. Objective was to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.03.6AR 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 3.03.6AR available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  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. Comparison of anti-neutrino reactor spectrum models with the Bugey 3 measurements

    NASA Astrophysics Data System (ADS)

    Achkar, B.; Aleksan, R.; Avenier, M.; Bagieu, G.; Bouchez, J.; Brissot, R.; Cavaignac, J. F.; Collot, J.; Cousinou, M.-C.; Cussonneau, J. P.; Declais, Y.; Dufour, Y.; Favier, J.; Garciaz, F.; Kajfasz, E.; de Kerret, H.; Koang, D. H.; Lefièvre, B.; Lesquoy, E.; Mallet, J.; Metref, A.; Nagy, E.; Obolensky, M.; Pessard, H.; Pierre, F.; Stutz, A.; Wuthrick, J. P.

    1996-02-01

    The Bugey 3 neutrino oscillation experiment has provided high statistics neutrino energy spectra recorded at 15 and 40 meters from a nuclear reactor core. Assuming no oscillations, the measured spectra favor a model of reactor spectrum based on the beta spectra measured at ILL.

  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. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  13. The future of reactor neutrino experiments: A novel approach to measuring theta{sub 13}

    SciTech Connect

    Heeger, Karsten M.; Freedman, Stuart J.; Luk, Kam-Biu

    2003-08-24

    Results from non-accelerator neutrino oscillation experiments have provided evidence for the oscillation of massive neutrinos. The subdominant oscillation, the coupling of the electron neutrino flavor to the third mass eigenstate, has not been measured yet. The size of this coupling U{sub e3} and its corresponding mixing angle theta{sub 13} are critical for CP violation searches in the lepton sector and will define the future of accelerator neutrino physics. The current best limit on U{sub e3} comes from the CHOOZ reactor neutrino disappearance experiment. In this talk we review proposals for future measurements of theta-13 with reactor antineutrinos.

  14. 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

  15. 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.

  16. Cryogenic Liquid Experiments in Orbit. Volume 2. Bubble Mechanics, Boiling Heat Transfer, and Propellant Tank Venting in a Zero-Gravity Environment

    DTIC Science & Technology

    1966-12-01

    Thermophoresis Experimental Apparatus . . . 27 8 Bubble Thermophoresis Lighting and Viewing Schematic ....... ..................... .... 28 9 Bubble... Thermophoresis Experimental Apparatus (with- out bubble holder) ...... ................ ... 29 10 Bubble Force Experiment Bubble Insertion Technique (Expanded View...several interesting and important ef- fects in bubble behavior and boiling heat transfer. These ef- fects are discussed below. 2. Bubble Thermophoresis

  17. Cosmogenic Neutron Production at the Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Mitchell, I.; Daya Bay Collaboration

    2017-09-01

    Neutrons are an important background for underground experiments studying neutrino oscillations, neutrino-less double-beta decay, dark matter, and other rare-event signals. The poster will present the status of a study of neutron production by cosmogenic muons at the Daya Bay Reactor Neutrino Experiment. The experiments configuration of multiple identical detectors at varying depths gives us the ability to measure neutron yield for different values of average muon energy within the same experiment. The current status of our study and future prospects will be discussed.

  18. 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)

  19. Safety Analysis of Small Break Loss of Coolant Accident for 1200 MWe Simplified Boiling Water Reactor (SBWR-1200 BDLB)

    SciTech Connect

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

    2002-07-01

    The objective of this research is to assess the performance of the safety systems during small break loss of coolant accident (SBLOCA) transient in the full size SBWR. RELAP5/MOD3 was used to simulate the blow-down and long-term cooling responses of the various safety systems during the accident transient. An integral test for long-term cooling under low pressure was conducted in a scaled facility with the initial conditions given by the code simulation. The code applicability and the facility scalability were evaluated by the comparison between the test data and the code simulations. The scaling analysis has been done by the comparison of the prototype code predictions and the scaled-up test data with the proper scaling multiplications and time shifting. The good agreement between the major safety parameters has shown the applicability of the RELAP5/MOD3 code and the scalability of the facility for SBWR-1200 safety analysis applications. (authors)

  20. Fifteen years experience filtering N reactor gaseous wastes

    SciTech Connect

    Fowler, K.L.

    1980-01-14

    The N Reactor exhaust gas filtering system consists of roughing filters, particulate filters, and charcoal filters, in that order. The basic particulate and charcoal filters consist of 0.46 m/sup 3//s, 60 cm x 60 cm x 30 cm filter canisters installed four canisters wide by ten canisters high in removable filter frames. The roughing filter canisters are about twice as wide and high as the other canisters but fit into similar frames of the same size. There are two side by side frames of each type of filter in each cell for a total design flow of 37 m/sup 3//s. The reactor and primary pipe space air has two cells for normal operation and one cell reserved for the accident case. One cell without roughing filters is used for the first buffer zone air around the reactor.

  1. 77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-22

    ... COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor... the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project Units 3 and 4 (STP 3 and...

  2. Experiment operations plan for the MT-4 experiment in the NRU reactor. [PWR

    SciTech Connect

    Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

    1983-06-01

    A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700/sup 0/F).

  3. Analysis of reactor material experiments investigating corium crust stability and heat transfer in jet impingement flow

    SciTech Connect

    Sienicki, J.J.; Spencer, B.W.

    1985-01-01

    Presented is an analysis of the results of the CSTI-1, CSTI-3, and CWTI-11 reactor material experiments in which a jet of molten corium initially at 3080/sup 0/K was directed downward upon a stainless steel plate. The experiments are a continuation of a program of reactor material tests investigating LWR severe accident phenomena. Objective of the present analysis is to determine the existence or nonexistence of a corium crust during impingement from comparison of the measured heatup of the plate (as measured by thermocouples imbedded immediately beneath the steel surface) with model calculations assuming alternately the presence and absence of a stable crust during impingement.

  4. 77 FR 76089 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-26

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  5. Results from the Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Tsang, K. V.; An, F. P.; An, Q.; Bai, J. Z.; Balantekin, A. B.; Band, H. R.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. L.; Cao, G. F.; Cao, J.; Carr, R.; Chan, W. T.; Chang, J. F.; Chang, Y.; Chasman, C.; Chen, H. S.; Chen, H. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, X. S.; Chen, Y.; Chen, Y. X.; Cherwinka, J. J.; Chu, M. C.; Cummings, J. P.; Deng, Z. Y.; Ding, Y. Y.; Diwan, M. V.; Draeger, E.; Du, X. F.; Dwyer, D.; Edwards, W. R.; Ely, S. R.; Fang, S. D.; Fu, J. Y.; Fu, Z. W.; Ge, L. Q.; Gill, R. L.; Gonchar, M.; Gong, G. H.; Gong, H.; Gornushkin, Y. A.; Gu, W. Q.; Guan, M. Y.; Guo, X. H.; Hackenburg, R. W.; Hahn, R. L.; Hans, S.; Hao, H. F.; He, M.; He, Q.; Heeger, K. M.; Heng, Y. K.; Hinrichs, P.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Huang, H. X.; Huang, H. Z.; Huang, X. T.; Huber, P.; Issakov, V.; Isvan, Z.; Jaffe, D. E.; Jetter, S.; Ji, X. L.; Ji, X. P.; Jiang, H. J.; Jiao, J. B.; Johnson, R. A.; Kang, L.; Kettell, S. H.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, C. Y.; Lai, W. C.; Lai, W. H.; Lau, K.; Lebanowski, L.; Lee, J.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Leung, K. Y.; Lewis, C. A.; Li, F.; Li, G. S.; Li, Q. J.; Li, W. D.; Li, X. B.; Li, X. N.; Li, X. Q.; Li, Y.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S. K.; Lin, Y. C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, J. C.; Liu, J. L.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Luk, A.; 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.; Mohapatra, D.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Nemchenok, I.; Ngai, H. Y.; Ngai, W. K.; Nie, Y. B.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevski, A.; Patton, S.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Shao, B. B.; Shih, K.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tagg, N.; Tam, Y. H.; Tanaka, H. K.; Tang, X.; Themann, H.; Torun, Y.; Trentalange, S.; Tsai, O.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viren, B.; 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, 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.; Williamson, Y.; Wise, T.; Wong, H. L. H.; Worcester, E. T.; Wu, F. F.; Wu, Q.; Xi, J. B.; Xia, D. M.; Xing, Z. Z.; Xu, J.; Xu, J.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, L.; Ye, M.; Yeh, M.; Yeh, Y. S.; Young, B. L.; Yu, Z. Y.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, S. H.; Zhang, Y. C.; Zhang, Y. H.; Zhang, Y. X.; Zhang, Z. J.; Zhang, Z. P.; Zhang, Z. Y.; Zhao, J.; Zhao, Q. W.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, Z. Y.; Zhuang, H. L.; Zou, J. H.

    2014-01-01

    The Daya Bay Reactor Neutrino Experiment was designed to achieve a sensitivity on the value of sin2 2θ13 to better than 0.01 at 90% CL. The experiment consists of eight antineutrino detectors installed underground at different baselines from six nuclear reactors. With data collected with six antineutrino detectors for 55 days, Daya Bay announced the discovery of a non-zero value for sin2 2θ13 with a significance of 5.2 standard deviations in March 2012. The most recent analysis with 139 days of data acquired in a six-detector configuration yields sin2 2θ13 = 0.089 ± 0.010 (stat.) ± 0.005 (syst.), which is the most precise measurement of sin2 2θ13 to date.

  6. The myth of the boiling point.

    PubMed

    Chang, Hasok

    2008-01-01

    Around 1800, many reputable scientists reported significant variations in the temperature of pure water boiling under normal atmospheric pressure. The reported variations included a difference of over 1 degree C between boiling in metallic and glass vessels (Gay-Lussac), and "superheating" up to 112 degrees C on extracting dissolved air out of water (De Luc). I have confirmed most of these observations in my own experiments, many of which are described in this paper. Water boils at the "boiling point" only under very particular circumstances. Our common-sense intuition about the fixedness of the boiling point is only sustained by our limited experience.

  7. 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.

  8. A review of film boiling at cryogenic temperatures.

    NASA Technical Reports Server (NTRS)

    Hsu, Y. Y.

    1972-01-01

    Film boiling occurs in the quenching of metals, the chilling of biological species, the regenerative cooling of rockets, and the cooling down of a cryogenic fuel tank. Occasionally film boiling is also found in a nuclear reactor or in a cryomagnet. Aspects of film boiling involving an unconstrained liquid mass are considered, giving attention to the evaporation time, the Leidenfrost temperature, solid-liquid contacts, the thermal properties of the solid, effects of coating or scale, wettability, the metastable condition, and the velocity effect on drops. Developments discussed with regard to pool boiling are related to vertical surfaces, film boiling from horizontal surfaces, film boiling from a horizontal cylinder, film boiling from a sphere, and film boiling of helium. Processes of film boiling in a channel are also analyzed.

  9. Glucose and 2,3-butanediol-paired syntheses in undivided flow reactors

    SciTech Connect

    Park, K.

    1985-01-01

    Paired synthesis, an energy-efficient electrochemical process, was applied to two biomass derived feedstocks. Sorbitol and calcium gluconate were produced by pairing electrochemical reduction of glucose with indirect electrogenerated bromine oxidation of glucose. The optimum electrode materials and operating conditions for the paired synthesis are: Raney nickel powder cathode, a packed bed anode of graphite chips, an initial glucose concentration of 1.6 M, a 0.4 M calcium bromide electrolyte, ca. pH 7, 60/sup 0/C, applied current of 25 mA per gram of cathode material, a solution velocity of 0.08 cm/sec, a six-minute residence time outside the reactor and parallel electrolyte solution and current flow. The 2,3-butanediol paired synthesis consisted of oxidation of acetoin by electrogenerated bromine at the anode followed by the electro-reduction of acetoin to 2-butanone at the cathode. The high boiling 2,3-butanediol is converted to the low boiling 2-butanone via the paired synthesis, which facilitates the recovery process. The optimum electrode materials and operating conditions for the 2,3-butanediol paired synthesis are: an amalgamated zinc cathode, a packed bed anode of graphite chips, 1 M 2,3-butanediol, 0.02 M acetoin, 1 M NaBr, ca. pH 7, 20/sup 0/C, 1.4 mA/cm/sup 2/ cathode current density, 1 mA/cm/sup 2/ anode current density, 0.04 cm/sec solution velocity, 4 minute residence time outside the reactor, and parallel electrolyte solution and current flow.

  10. 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.

  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. Light water reactor mixed-oxide fuel irradiation experiment

    SciTech Connect

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-06-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding.

  13. 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.

  14. 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.

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

    DOE PAGES

    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

  16. 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.

  17. 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.

  18. 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.

  19. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    SciTech Connect

    Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  20. Verification and Validation of Neutronic/Thermalhydraulic 3D-Time Dependent Model for Treatment of Super-critical States of Light water Research Reactors Accidents

    SciTech Connect

    Khaled, S.M.

    2015-07-01

    This work presents the Verification and testing both the neutronic and thermal-hydraulics response of the positive reactivity-initiated power excursion accidents in small light water research reactors. Some research reactors have to build its own severe accidents code system. In this sense, a 3D space-time-dependent neutron diffusion models with thermal hydraulic feedback have been introduced, compared and tested both experimentally at criticality 14-cent and theoretically up to 1.5 $ with a number of similar codes. The results shows that no expected core failure or moderator boiling. (author)

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Thermodynamic analysis of helium boil-off experiments with pressure variations

    NASA Astrophysics Data System (ADS)

    Cha, Y. S.; Niemann, R. C.; Hull, J. R.

    A thermodynamic analysis by calorimetric experiments in a system with changing pressure is presented. A general equation is derived for use in calculating the rate of heat loss from measured mass flow rate. The results show that the largest contribution from pressure variation is the sensible heat of liquid helium in a Dewar. A dimensionless parameter that was identified provides an indication of the importance of pressure variation relative to the latent heat of vaporization during an experiment. This dimensionless parameter is a function of system pressure land the thermodynamic properties of helium), rate of change of system pressure, liquid helium inventory in the Dewar and measured mass flow rate. In the special case when the effect of pressure variation is small compared to the latent heat of vaporization, the heat loss rate is the product of the measured mass flow rate and the latent heat of vaporization, multiplied by a correction factor that is a function of the ratio of vapour density to liquid density. This correction factor is significant for helium at pressures near or above 1 atm and should always be included in the calculation.

  7. 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.

  8. 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.

  9. Results of small break LOCA experiments in the LOFT reactor system with comparison to code calculations. [PWR

    SciTech Connect

    Adams, J.P.; Linebarger, J.H.; Leach, L.P.

    1980-01-01

    The results are presented of three small break loss-of-coolant experiments performed in the LOFT Pressurized Water Reactor (PWR) system. Experiment L3-0, performed without reactor power, represented a loss of coolant from the power operated relief valve on the top of the pressurizer. Experiments L3-1 and L3-2 were initiated with the reactor at full power (maximum linear heat generation rate approximately 52 kW/m) and represented 4-in and 1-in diameter breaks, respectively, in the reactor inlet piping of a commercial PWR. Comparisons of data to analytical model calculations with a number of different models indicate that most major phenomena were correctly calculated, but that improvements in modeling small break behavior are necessary.

  10. 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.

  11. 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.

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

    DOE PAGES

    An, F. P.; Hackenburg, R. W.; Brown, R. E.; ...

    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)

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

    SciTech Connect

    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.; Kettell, S. H.; Littenberg, L.; Pearson, C. E.; Qian, X.; Theman, H.; Viren, B.; Worcester, E.; Yeh, M.; Zhang, C.

    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)

  14. 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.

  15. New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

    NASA Astrophysics Data System (ADS)

    Ma, X. B.; Qiu, R. M.; Chen, Y. X.

    2017-02-01

    Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between 235U and 239Pu, the covariance coefficient changes from 0.15 to -0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller.

  16. Summary of the fourth conference on United States utility experience in reactor noise analysis

    SciTech Connect

    Fry, D.N.

    1987-01-01

    The fourth informal conference on United States utility experience in reactor noise analysis and loose-part monitoring was held at the Northeast Utilities Service Company offices in Hartford, Connecticut, May 12-14, 1987. Host and general chairman for the meeting was J.V. Persio of Northeast Utilities. This conference provided a forum where utilities could share information on reactor noise analysis on an informal basis. There were about 60 attendees at the meeting representing 10 utilities, 3 reactor vendors, 8 consulting organizations, and 4 universities and research laboratories. Twenty-three papers were presented at the conference, dealing with various aspects of loose-part monitoring, neutron noise analysis, and standards activities.

  17. 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.

  18. 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.

  19. Management of historical waste from research reactors: the Dutch experience

    SciTech Connect

    Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron

    2013-07-01

    Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

  20. 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.

  1. 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.

  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. New instruments at the upgraded JRR-3 reactor

    NASA Astrophysics Data System (ADS)

    Funahashi, Satoru

    1991-10-01

    A new JRR-3 reactor equipped with a liquid hydrogen cold source started to operate at 20 MW full power for general users in November 1990. The reactor provides seven horizontal beam holes in the reactor room and three cold guides and two thermal ones in the guide hall. In total about twenty five instruments are planned for use with the neutron beam. Eight triple-axis neutron spectrometers, two small-angle scattering machines, two high-resolution powder diffractometers, two radiography instruments are included among these. More than half of them have already been completed and the rest will be installed in the next two years.

  4. 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.

  5. Identification and Ranking of Phenomena Leading to Peak Cladding Temperatures in Boiling Water Reactors During Large Break Loss of Coolant Accident Transients

    SciTech Connect

    Ratnayake, Ruwan K.; Ergun, S.; Hochreiter, L.E.; Baratta, A.J.

    2002-07-01

    In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. The establishment of Phenomena Identification and Ranking Tables (PIRT) provides a vehicle to assist in assessing the capabilities of the computer code, and to guide the uncertainty analysis of the calculated results. The process used in this work to identify the phenomena was reviewing both licensing and best estimate calculations, as well as experiments, which had been performed for BWR LOCA analyses. The initial PIRT was developed by a group of analysts and was compared to existing BWR LOCA PIRTs as well as BWR LOCA analyses. The initial PIRT was then independently reviewed by a second panel of experts for the selected ranking of phenomena, identification of phenomena which were ignored, as well as the basis and rationale for the ranking of the phenomena. The differences between the two groups were then resolved. PIRTs have been developed for BWR types 4 and 5/6 for the Large Break Loss of Coolant Accidents (LB-LOCA). The ranking and the corresponding rationale for each phenomenon is included in tables together with the assessed uncertainty of the code capability to predict the phenomena. (authors)

  6. Reactor Dynamics Experiments with a Sub-Critical Assembly

    SciTech Connect

    Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

    2004-10-06

    A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory.

  7. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    NASA Astrophysics Data System (ADS)

    Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

    2017-09-01

    VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

  8. Lessons learned from applying VIM to fast reactor critical experiments

    SciTech Connect

    Schaefer, R.W.; McKnight, R.D.; Collins, P.J.

    1995-05-17

    VIM is a continuous energy Monte Carlo code first developed around 1970 for the analysis of plate-type, fast-neutron, zero-power critical assemblies. In most respects, VIM is functionally equivalent to the MCNP code but it has two features that make uniquely suited to the analysis of fast reactor critical experiments: (1) the plate lattice geometry option, which allows efficient description of and neutron tracking in the assembly geometry, and (2) a statistical treatment of neutron cross section data in the unresolved resonance range. Since its inception, VIM`s capabilities have expanded to include numerous features, such as thermal neutron cross sections, photon cross sections, and combinatorial and other geometry options, that have allowed its use in a wide range of neutral-particle transport problems. The earliest validation work at Argonne National Laboratory (ANL) focused on the validation of VIM itself. This work showed that, in order for VIM to be a ``rigorous`` tool, extreme detail in the pointwise Monte Carlo libraries was needed, and the required detail was added. The emphasis soon shifted to validating models, methods, data and codes against VIM. Most of this work was done in the context of analyzing critical experiments in zero power reactor (ZPR) assemblies. The purpose of this paper is to present some of the lessons learned from using VIM in ZPR analysis work. This involves such areas as uncovering problems in deterministic methods and models, pitfalls in using Monte Carlo codes, and improving predictions. The numerical illustrations included here were taken from the extensive documentation cited as references.

  9. Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

    NASA Astrophysics Data System (ADS)

    Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

    2017-07-01

    In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

  10. The effects of geometric, flow, and boiling parameters on bubble growth and behavior in subcooled flow boiling

    NASA Astrophysics Data System (ADS)

    Samaroo, Randy

    Air bubble injection and subcooled flow boiling experiments have been performed to investigate the liquid flow field and bubble nucleation, growth, and departure, in part to contribute to the DOE Nuclear HUB project, Consortium for Advanced Simulation of Light Water Reactors (CASL). The main objective was to obtain quantitative data and compartmentalize the many different interconnected aspects of the boiling process -- from the channel geometry, to liquid and gas interactions, to underlying heat transfer mechanisms. The air bubble injection experiments were performed in annular and rectangular geometries and yielded data on bubble formation and departure from a small hole on the inner tube surface, subsequent motion and deformation of the detached bubbles, and interactions with laminar or turbulent water flow. Instantaneous and ensemble- average liquid velocity profiles have been obtained using a Particle Image Velocimetry technique and a high speed video camera. Reynolds numbers for these works ranged from 1,300 to 7,700. Boiling experiments have been performed with subcooled water at atmospheric pres- sure in the same annular channel geometry as the air injection experiments. A second flow loop with a slightly larger annular channel was constructed to perform further boiling experiments at elevated pressures up to 10 bar. High speed video and PIV measurements of turbulent velocity profiles in the presence of small vapor bubbles on the heated rod are presented. The liquid Reynolds number for this set of experiments ranged from 5,460 to 86,000. It was observed that as the vapor bubbles are very small compared to the injected air bubbles, further experiments were performed using a microscopic objective to obtain higher spatial resolution for velocity fields near the heated wall. Multiple correlations for the bubble liftoff diameter, liftoff time and bub- ble history number were evaluated against a number of experimental datasets from previous works, resulting in a

  11. FELIX experiments and computational needs for eddy current analysis of fusion reactors

    SciTech Connect

    Turner, L.R.

    1984-01-01

    In a fusion reactor, changing magnetic fields are closely coupled to the electrically-conducting metal structure. This coupling is particularly pronounced in a tokamak reactor in which magnetic fields are used to confine, stabilize, drive, and heat the plasma. Electromagnetic effects in future fusion reactors will have far-reaching implications in the configuration, operation, and maintenance of the reactors. This paper describes the impact of eddy-current effects on future reactors, the requirements of computer codes for analyzing those effects, and the FELIX experiments which will provide needed data for code validation.

  12. Determination of the neutron flux for the Yankee Rowe experiment in the Ford Nuclear Reactor

    SciTech Connect

    Cacciapouti, R.J.; Petrusha, L.

    1994-12-31

    Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented.

  13. A review of experiments and results from the transient reactor test (TREAT) facility.

    SciTech Connect

    Deitrich, L. W.

    1998-07-28

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

  14. A search for neutrino oscillations using the CHOOZ 1 km baseline reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    George, Jean

    1999-10-01

    Neutrino oscillation searches are an active field of research due to the implications their discovery may have for the solar neutrino anomaly as well as for the atmospheric neutrino anomaly. Their discovery may also have broad ramifications for the Standard Model of Particle Physics as a whole. Results from an oscillation search using the CHOOZ long baseline reactor neutrino experiment are presented in this thesis. These results are based on the data taken from June 1997 through April 1998 when the two reactors ran at combined thermal power levels ranging from zero power to their full power level of 8.5 GW. Electron flavored antineutrinos emanating from the reactors were detected through the inverse beta decay channel using a liquid scintillating calorimeter located at a distance of approximately 1 km from the reactor sources. The underground experimental site (300 MWE) provided natural shielding from the background of cosmic ray muons-leading to a background rate more than an order of magnitude lower than the full power signal rate. From the agreement between the detected and expected neutrino event rates no evidence for neutrino oscillations was found (at the 90% C.L.) for the oscillation parameter space governed by Δm 2 > 0.8 × 10-3 eV2 for maximal mixing and by sin2 2Θ > 0.18 for large values of Δm2.

  15. Review and evaluation of the RELAP5YA computer code and the Vermont Yankee LOCA (Loss-of-Coolant Accident) licensing analysis model for use in small and large break BWR (Boiling Water Reactor) LOCAS

    SciTech Connect

    Jones, J.L.

    1987-01-01

    A review has been completed of the RELAP5YA computer code to determine its acceptability for performing licensing analyses. The review was limited to Boiling Water Reactor (BWR) reactor applications. In addition, a Loss-Of-Coolant Accident (LOCA) licensing analysis method, using the RELAP5YA computer code, has been reviewed. This method is applicable to the Vermont Yankee Nuclear Power Station to perform full break spectra LOCA and fuel cycle independent analyses. The review of the RELAP5YA code consisted of an evaluation of all Yankee Atomic Electric Company (YAEC) incorporated modifications to the RELAP5/MOD1 Cycle 18 computer code from which the licensing version of the code originated. Qualifying separate and integral effects assessment calculations were reviewed to evaluate the validity and proper implementation of the various added models. The LOCA licensing method was assessed by reviewing two RELAP5YA system input models and evaluating several small and large break qualifying transient calculations. A review of the RELAP5YA code modifications and their assessments, as well as the submitted LOCA licensing method, is given and the results of the review are provided.

  16. 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)

  17. 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.

  18. Constraints on very light sterile neutrinos from θ 13-sensitive reactor experiments

    NASA Astrophysics Data System (ADS)

    Palazzo, Antonio

    2013-10-01

    Three dedicated reactor experiments, Double Chooz, RENO and Daya Bay, have recently performed a precision measurement of the third standard mixing angle θ 13 exploiting a multiple baseline comparison of ν e → ν e disappearance driven by the atmospheric mass-squared splitting. In this paper we show how the same technique can be used to put stringent limits on the oscillations of the electron neutrino into a fourth very light sterile species (VLS ν) characterized by a mass-squared difference lying in the range [10-3 - 10-1] eV2. We present accurate constraints on the admixture | U e4|2 obtained by a 4-flavor analysis of the publicly available reactor data. In addition, we show that the estimate of θ 13 obtained by the combination of the three reactor experiments is rather robust and substantially independent of the 4-flavor-induced perturbations provided that the new mass-squared splitting is not too low (≳ 6 × 10-3 eV2). We briefly comment on the possible impact of VLS ν's on the rest of the neutrino oscillation phenomenology and emphasize their potential role in the cosmological "dark radiation" anomaly.

  19. Reactor ν̄e disappearance in the Double Chooz experiment

    DOE PAGES

    Abe, Y.; Aberle, C.; dos Anjos, J. C.; ...

    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).

  20. 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).

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Lessons Learned about Liquid Metal Reactors from FFTF Experience

    SciTech Connect

    Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.; Burke, Thomas M.; Grandy, Christopher

    2016-09-20

    The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens. In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports

  8. 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.

  9. Recent BeO-reflector-controlled reactor experiments in ZPPR

    SciTech Connect

    McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.; McKnight, R.D.

    1986-01-01

    Integral reactor physics measurements were performed on a BeO-reflected fast reactor assembly in the ZPPR facility during January and February of 1985. The measurements emphasized power distributions and reflector control worths in two different critical states. The measurements have been analyzed using three-dimensional deterministic and Monte Carlo methods and the ENDF/B-V.2 nuclear data library. Together the measurements and analyses form a modern, reliable, benchmark data set for testing calculational methods that will be used in predicting some of the design parameters for future space reactors.

  10. 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.

  11. 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.

  12. 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)

  13. 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.

  14. Calculation to experiment comparison of SPND signals in various nuclear reactor environments

    SciTech Connect

    Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien; Snoj, Luka; Tarchalski, Mikolaj; Dewynter-Marty, Veronique; Malouch, Fadhel

    2015-07-01

    In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)

  15. The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

    NASA Astrophysics Data System (ADS)

    Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

    The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

  16. 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

  17. Comments on the determination of the neutrino mass ordering in reactor neutrino experiments

    NASA Astrophysics Data System (ADS)

    Bilenky, S. M.

    2017-05-01

    We consider the problem of determination of the neutrino mass ordering via precise study of the vacuum neutrino oscillations in the JUNO and other future medium baseline reactor neutrino experiments. We are proposing to resolve neutrino mass ordering by determination of the neutrino oscillation parameters from analysis of the data of the reactor experiments and comparison them with the oscillation parameters obtained from analysis of the solar and KamLAND experiments.

  18. Little low-power boiling never hurt anybody. [LMFBR

    SciTech Connect

    Dunn, F.E.

    1985-01-01

    Failures in the shutdown heat removal system of an LMFBR might lead to flow stagnation and coolant boiling in the reactor core. At normal operating power, the onset of sodium boiling will lead to film dryout and melting of the cladding and fuel within a few seconds. On the other hand, both calculations and currently available experimental data indicate that at heat fluxes corresponding to decay heat power levels, boiling leads to improved heat removal; and it limits the temperature rise in the fuel pins. Therefore, when setting safety criteria for decay heat removal systems, there is no reason to preclude sodium boiling per se because of heat removal considerations. As an example that illustrates the beneficial impact of coolant boiling, a case involving temporary loss of feedwater and staggered pump failures in a hypothetical, 1000-MWe loop-type reactor was run in the SASSYS-1 code.

  19. Boiling heat transfer enhancement in subsurface horizontal and vertical tunnels

    SciTech Connect

    Pastuszko, Robert

    2008-09-15

    Complex experimental investigations of boiling heat transfer on structured surfaces covered with perforated foil were taken up. Experimental data were discussed for two kinds of enhanced surfaces formed by joined horizontal and vertical tunnels: tunnel structures (TS) and narrow tunnel structures (NTS). The experiments were carried out with water, ethanol and R-123 at atmospheric pressure. The TS and NTS surfaces were manufactured out of perforated copper foil of 0.05 mm thickness (hole diameters: 0.3, 0.4, 0.5 mm) sintered with the mini-fins, formed on the vertical side of the 5 mm high rectangular fins and horizontal inter-fin surface. The effects of hole (pore) diameters, tunnel pitch for TS and tunnel width for NTS on nucleate pool boiling were examined. Substantial enhancement of heat transfer coefficient was observed. The investigated surfaces showed boiling heat transfer coefficients similar to those of existing structures with subsurface tunnels, but at higher heat fluxes range. (author)

  20. Colloidal 3-mercaptopropionic acid Capped Lead Sulfide Quantum Dots in a Low Boiling Point Solvent.

    PubMed

    Reinhart, Chase C; Johansson, Erik

    2017-04-10

    Colloidal 3-mercaptopropionic acid (3-MPA) capped lead sulfide quantum dots were prepared in a variety of organic solvents stabilized with a quaternary ammonium halide salt. The stabilized colloids' optical properties were studied through optical absorption and emission spectroscopy and found to be dependent on both the concentration of new ligand and stabilizer, and sample age. Nanocrystal ligand chemistry was studied through a combination of (1)H-NMR and 2-dimensional Nuclear Overhauser Effect Spectroscopy (NOESY) which revealed full displacement of the original oleate ligand to form a dynamically exchanging ligand shell. The colloids were studied optically and via NMR as they aged and revealed a quantitative conversion of monomeric 3-mercaptopropionic acid to its dimer, dithiodipropionic acid (dTdPA).

  1. 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.

  2. 3D calculation of boiling in complex geometry of steam generator

    SciTech Connect

    Ravnikar, I.; Petelin, S.

    1996-11-01

    The D4 steam generator of nuclear power plant Krsko was analyzed from a thermal-hydraulic point of view using the 3D PHOENICS computer code. The calculations were carried out for different operating conditions based on plugging study that was performed using 1D steam generator model SMUP computer code. Void fraction, velocity and enthalpy distributions were then obtained in the U-tube riser section.

  3. Irradiation creep of VTiCr alloy in BR-10 reactor core instrumented experiments

    NASA Astrophysics Data System (ADS)

    Troyanov, V. M.; Bulkanov, M. G.; Kruglov, A. S.; Krjuchkov, E. A.; Nikulin, M. P.; Pevchykh, J. M.; Rusanov, A. E.; Smirnoff, A. A.; Votinov, S. N.

    1996-10-01

    A thin wall tubular-type speciment of 4%Ti-4%Cr vanadium alloy was tested for creep under irradiation in BR-10 reactor at 713-723 K and at 8.6 × 10 18 n/m 2s fast neutron flux. A fluence at the end of the experiment have reached 5.8 × 10 25 n/m 2. Specimen deformation measurements were performed by a dynamometric method based on a stress relaxation control provided during irradiation under constant load applied. During the experiment 13 deformation curves were obtained for different stress levels ranged up to 165 MPa. At the same time the yield stress of the irradiated specimen was periodically determined. The irradiation creep rate has been found to be proportional to the stress up to 110-120 MPa with the module equal to 3.3 × 10 -12 dpa -1Pa -1. At higher streses, a creep process essentially accelerates. The results on VTiCr alloy are discussed in respect to data obtained for stainless steels in earlier BR-10 reactor experiments.

  4. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - main report. Final report

    SciTech Connect

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2), which is a boiling water reactor (BWR), located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low- level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.

  5. 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

  6. An alternative method of determining the neutrino mass ordering in reactor neutrino experiments

    NASA Astrophysics Data System (ADS)

    Bilenky, S. M.; Capozzi, F.; Petcov, S. T.

    2017-09-01

    We discuss a novel alternative method of determining the neutrino mass ordering in medium baseline experiments with reactor antineutrinos. Results on the potential sensitivity of the new method are also presented.

  7. Characteristics of releases from TREAT source term experiment STEP-3

    SciTech Connect

    Fink, J.K.; Schlenger, B.J.; Baker, L. Jr.; Ritzman, R.L.

    1987-01-01

    Four in-pile experiments designed to characterize the radiological source term associated with postulated severe light water reactor accidents were performed at the Transient Reactor Test Facility. STEP-3 simulated a high-pressure TMLB' pressurized water reactor accident sequence that includes the extended loss of all ac power and leads to the loss of long-term decay heat removal. In STEP-3, four fuel elements from the Belgonucleaire BR3 reactor were subjected to temperature and pressures approaching those of a TMLB' accident. A description of the experiment and thermal-hydraulic analysis is reported elsewhere. The aerosols released into the flow stream were collected on coupons, settling plates, and wire impactors. Examination of the collected aerosol deposits was performed using scanning electron microscopy, electron microprobe microanalysis, and secondary ion mass spectroscopy (SIMS), to provide information about the chemical composition and morphology of the release. This paper describes the aerosol deposits and elemental composition of the release.

  8. Boiling rocks

    SciTech Connect

    Not Available

    1992-09-01

    This paper discusses Zeolites, 3-dimensional crystals made up to silicon, aluminum, oxygen, and small amounts of other elements. Highly porous, zeolites differ in crystallization and composition. However, both natural and synthetic zeolites, of which in there are about 133, are characterized by submicroscopic channels and holes, often called pores, that let zeolites act as molecular sieves. It is this molecular-sieve capability that has made zeolites so valuable as a catalyst in industrial uses. According to a leading zeolite authority, Dr. John M. Newsam, a director with BIOSYM Technologies, Inc., Every crude oil developer worldwide uses a zeolite as a catalyst. So it's big business. Besides their use in reducing the cost of processing gasoline and other petroleum products, zeolites are helping in cleaning up low-level nuclear wastes and other hazardous materials. They're also used in aromatic processing and in raising pigs and tomatoes. In the coming years, zeolites will used in place of phosphorous in certain products.

  9. Analysis and evaluation of ZPPR (Zero Power Physics Reactor) critical experiments for a 100 kilowatt-electric space reactor

    SciTech Connect

    McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W. ); Doncals, R.A.; Andre, S.V.; Porter, C.A. ); Cowan, C.L; Stewart, S.L.; Protsik, R. . Astro Space Div.)

    1990-01-01

    ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously applied to fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further optimization. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design. 13 refs., 5 figs., 7 tabs.

  10. 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.

  11. Introduction to D-He(3) fusion reactors

    NASA Astrophysics Data System (ADS)

    Vlases, G. C.; Steinhauer, L. C.

    1989-07-01

    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.

  12. 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

  13. 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.

  14. Integral experiment information for fast reactors: Sensitivity and uncertainty analysis of reactor performance parameters

    SciTech Connect

    Collins, P.J.

    1982-01-01

    This chapter offers a detailed analysis of uncertainties in experimental parameters for the ZPR benchmark cores. Discusses the critical facilities and measurements; the need for well documented data; the relevance of data for reactor design; uses of integral data; benchmark data; mockup cores; accuracy of experimental data; critical mass; reaction rate ratios; covariance matrices; selection of reliable integral data; cavity measurements; and the SCHERZO 556 core. Points out that substantial revisions of data in the CSEWG benchmark book have resulted from a reevaluation of analytical corrections using modern methods and codes. Concludes that the integral data presently being utilized represent a very limited base, which will be enlarged considerably before application to a wider range of power reactor parameters.

  15. Leukemia in the proximity of a German boiling-water nuclear reactor: evidence of population exposure by chromosome studies and environmental radioactivity.

    PubMed Central

    Schmitz-Feuerhake, I; Dannheim, B; Heimers, A; Oberheitmann, B; Schröder, H; Ziggel, H

    1997-01-01

    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

  16. 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

  17. 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.

  18. RELAP5 Simulation of Thermal-Hydraulic Behavior in a CANDU Reactor - Assessments of RD-14 Experiments

    SciTech Connect

    Lee, Sukho; Kim, In-Goo

    2000-04-15

    The critical reactor header break and the thermosiphoning experiments in the RD-14 test facility were simulated with the RELAP5/MOD3.1 code. The RELAP5 code has been developed for best-estimate transient simulation of pressurized water reactors and associated systems, but it has not been assessed for a Canada deuterium uranium (CANDU) reactor. Therefore, this study has been initiated with an aim to identify the code applicability in a CANDU reactor by simulating some of the tests performed in the RD-14 facility. The RD-14 test facility at Whiteshell Nuclear Research Establishment is a full-scale pressurized-water loop. The RD-14 is not a scale model of any particular CANDU reactor. Rather, it possesses many geometric features of a CANDU reactor heat transport system and is capable of operating at conditions similar to those expected to occur in a reactor under normal operation and some postulated accident conditions. In this study, two critical reactor header break tests (B8711 and B8713) and three thermosiphoning tests (T8513, T8515, and T8517) were analyzed with the RELAP5 code. The results were compared with experimental data and those of CATHENA performed by Atomic Energy of Canada Ltd. The RELAP5 analyses demonstrate the code's capability to predict reasonably the main phenomena occurring in the transient, in both the qualitative and the quantitative view. However, some discrepancies after the emergency coolant injection for the critical break case and also related to the behaviors of the mass flow rate and the primary pressure for the thermosiphoning case were observed.

  19. 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.

  20. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    NASA Astrophysics Data System (ADS)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  1. Decommissioning of the BR3 reactor: status and perspectives

    SciTech Connect

    Noynaert, L.; Verstraeten, I.

    2007-07-01

    The BR3 plant at Mol in Belgium built at the end of the fifties was the first PWR plant built outside the USA. The reactor had a small net power output (10 MWe) but comprised all the loops and features of a commercial PWR plant. The BR3 plant was operated with the main objective of testing advanced PWR fuels under irradiation conditions similar to those encountered in large commercial PWR plants. The reactor was started in 1962 and shut down in 1987 after 25 years of continuous operation. Since 1989, SCK.CEN is decommissioning the BR3 PWR research reactor. The dismantling of the metallic components including reactor pressure vessel and internals is completed and extensively reported in the literature. The dismantling of auxiliary components and the decontamination of parts of the infrastructure are now going on. The decommissioning progress is continuously monitored and costs and strategy are regularly reassessed. The first part of the paper describes the main results and lessons learned from the reassessment exercises performed in 1994, 1999, 2004 and 2007. Impacts of changes in legal framework on the decommissioning costs will be addressed. These changes concern e.g. licensing aspects, clearance levels, waste management... The middle part of the paper discusses the management of activated and/or contaminated concrete. The costing exercise performed in 1995 highlighted that the management of activated and contaminated concrete is the second main cost item after the dismantling of the reactor pressure vessel and internals. Different possible solutions were studied. These are evacuation as radioactive waste with or without supercompaction, recycling this 'radioactive' grout or concrete for conditioning of radioactive waste e.g. conditioning of metallic waste. The paper will give the results of the cost-benefit analysis made to select the solution retained. The last part of the paper will discuss the end goal of the decommissioning of the BR3. In the final

  2. 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

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

  3. Neutral beam injection in a D 3He FRC reactor

    NASA Astrophysics Data System (ADS)

    Ferrari, Hugo; Farengo, Ricardo

    2007-06-01

    The use of neutral beam injection (NBI) to sustain a fraction of the plasma current in a field reversed configuration (FRC) reactor operating with the D-3He reaction is studied. A Monte Carlo code already used to study NBI in medium size FRCs is employed (Lifschitz A F, Farengo R and Arista N R 2002 Nucl. Fusion 42 863, Lifschitz A F, Farengo R and Arista N R 2002 Plasma Phys. Control. Fusion 44 1979, Lifschitz A F, Farengo R and Hoffman A L 2004 Nucl. Fusion 44 1015) and the plasma parameters are similar to those proposed in the ARTEMIS (Momota H, Ishida A, Kohzaki Y, Miley G, Ohi S, Ohnishi M, Sato K, Steinhauer L, Tomita Y and Tuszewki M 1992 Fusion Technol. 21 2307) conceptual reactor design. A simple analysis shows that the driven current cannot reach the values quoted in the ARTEMIS project and a procedure to search for plasma parameters that result in higher efficiencies is presented.

  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. 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

  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. 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…

  9. 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…

  10. 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…

  11. 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.

  12. 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.

  13. 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.

  14. Overview of the FUTURIX-FTA Irradiation Experiment in the Phénix Reactor

    SciTech Connect

    Heather J.M. Chichester; Steve L. Hayes; Kenneth J. McClellan; Jean-Luc Paul; Marc Masson; Stewart L. Voit; Fabienne Delage

    2015-09-01

    The Advanced Fuels Campaign utilizes the Advanced Test Reactor (ATR) for most of its irradiation testing. Cadmium-shrouded baskets are used in ATR to modify the neutron spectrum to simulate a fast reactor environment for the fuel. FUTURIX-FTA is an irradiation experiment conducted in the Phenix fast reactor in France. Results from FUTURIX-FTA and irradiation tests in ATR using identical fuel compositions will be compared to identify and evaluate any differences in fuel behavior due to differences in the irradiation source.

  15. 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.

  16. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  17. 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.

  18. 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.

  19. 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.

  20. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    SciTech Connect

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.

  1. 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.

  2. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    SciTech Connect

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-07-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at {approx}2.4, {approx}7 and {approx}11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of {approx}7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10{sup 15} n/cm{sup 2}/s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between {approx}410 deg. C and {approx}645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  3. 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

  4. 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.

  5. Investigation of a Possibility of Chromium-51 Accumulation in the SM-3 Reactor to Fabricate a Neutrino Source

    NASA Astrophysics Data System (ADS)

    Romanov, E. G.; Gavrin, V. N.; Tarasov, V. A.; Malkov, A. P.; Kupriyanov, A. V.; Danshin, S. N.; Veretenkin, E. P.

    2017-01-01

    Compact high intensity neutrino sources based on 51Cr isotope are demanded for very short baseline neutrino experiments. In particular, a 3 MCi 51Cr neutrino source is needed for the experiment BEST on search for transitions of electron neutrinos to sterile states. The paper presents the results of the analysis of options of the irradiation of highly enriched 50Cr in the existing trap of thermal neutrons of high-flux reactor SM–3, as well as using the most promising variants of the trap after upcoming reconstruction of the reactor. It is shown that it is possible to to obtain the intensity of 51Cr up to 3.85 MCi at the end of irradiation of 50Cr enriched to 97% in the high-flux reactor SM–3 of the JSC “SSC NIIAR”.

  6. Study of thermal behavior of vitamin D3 by pyrolysis-GC-MS in combination with boiling point-retention time correlation.

    PubMed

    Sun, Yu'an; Liu, Baoxia; Wang, Guoqing; Zhang, Rongjie; Xie, Bing

    2005-01-01

    The thermal behavior of vitamin D3 was studied based on pyrolysis-GC-MS technique. It was pyrolyzed at 600 degrees C, 750 degrees C, 900 degrees C, respectively. The pyrolysis product were separated With an HP-5 column and identified by the NIST mass spectral search program in combination with the correlation of boiling point and retention time (BP-RT). There are totally 50 components, including mono aromatics and polycyclic aromatic hydrocarbons (PAHs), were determined. It is shown that the contents of the PAHs are increasing with the increasing of the pyrolysis temperature. The contents of the determined components vary from 0.04% to 37.08%.

  7. FAFTRCS: an experiment in computerized reactor safety systems

    SciTech Connect

    Chisholm, G.H.

    1985-01-01

    Nuclear Power Plant availability and reliability could be improved by the integration of computers into the control environment. However, computer-based systems are historically viewed as being unreliable. This places a burden upon the designer to demonstrate adequate reliability and availability for the computer. The complexity associated with computers coupled with the manual nature of these demonstrations results in a high cost which typically has been justified for critical applications only. This paper investigates a methodology for automating this process and discusses a project which intends to apply this methodology to design verification and validation for a control system which will be installed and tested in an actual reactor control environment. 7 refs., 4 figs., 1 tab.

  8. Thermosiphon-based PCR reactor: experiment and modeling.

    PubMed

    Chen, Zongyuan; Qian, Shizhi; Abrams, William R; Malamud, Daniel; Bau, Haim H

    2004-07-01

    A self-actuated, flow-cycling polymerase chain reaction (PCR) reactor that takes advantage of buoyancy forces to continuously circulate reagents in a closed loop through various thermal zones has been constructed, tested, and modeled. The heating required for the PCR is advantageously used to induce fluid motion without the need for a pump. Flow velocities on the order of millimeters per second are readily attainable. In our preliminary prototype, we measured a cross-sectionally averaged velocity of 2.5 mm/s and a cycle time of 104 s. The flow velocity is nearly independent of the loop's length, making the device readily scalable. Successful amplifications of 700- and 305-bp fragments of Bacillus cereus genomic DNA have been demonstrated. Since the device does not require any moving parts, it is particularly suitable for miniature systems.

  9. Operating experience feedback report: Experience with pump seals installed in reactor coolant pumps manufactured by Byron Jackson. Commercial power reactors, Volume 7

    SciTech Connect

    Bell, L.G.; O`Reilly, P.D.

    1992-09-01

    This report examines the reactor coolant pump (RCP) seal operating experience through August 1990 at plants with Byron Jackson (B-J) RCPs. ne operating experience examined in this analysis included a review of the practice of continuing operation with a degraded seal. Plants with B-J RCPs that have had relatively good experience with their RCP seals attribute this success to a combination of different factors, including: enhanced seal QA efforts, modified/new seal designs, improved maintenance procedures and training, attention to detail, improved seal operating procedures, knowledgeable personnel involved in seal maintenance and operation, reduction in frequency of transients that stress the seals, seal handling and installation equipment designed to the appropriate precision, and maintenance of a clean seal cooling water system. As more plants have implemented corrective measures such as these, the number of B-J RCP seal failures experienced has tended to decrease. This study included a review of the practice of continued operation with a degraded seal in the case of PWR plants with Byron Jackson reactor coolant pumps. Specific factors were identified which should be addressed in order to safety manage operation of a reactor coolant pump with indications of a degrading seal.

  10. 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.

  11. Boiling from small cylinders.

    NASA Technical Reports Server (NTRS)

    Bakhru, N.; Lienhard, J. H.

    1972-01-01

    Heat transfer is observed as a function of temperature on small horizontal wires in water and four organic liquids. When the wire radius is sufficiently small, the hydrodynamic transitions in the boiling curve disappear and the curve becomes monotonic. Three modes of heat removal are identified for the monotonic curve and described analytically: a natural convection mode, a mixed film boiling and natural convection mode, and a pure film boiling mode. Nucleate boiling does not occur on the small wires. The study was motivated by an interest in predicting the behavior of large heaters at low gravity. The application of the present results to such circumstances is therefore discussed. It is proposed that the peak and minimum heat fluxes will vanish at low gravity as well as on small wires.

  12. Experiment on search for neutron–antineutron oscillations using a projected UCN source at the WWR-M reactor

    NASA Astrophysics Data System (ADS)

    Fomin, A. K.; Serebrov, A. P.; Zherebtsov, O. M.; Leonova, E. N.; Chaikovskii, M. E.

    2017-01-01

    We propose an experiment on search for neutron–antineutron oscillations based on the storage of ultracold neutrons (UCN) in a material trap. The sensitivity of the experiment mostly depends on the trap size and the amount of UCN in it. In Petersburg Nuclear Physics Institute (PNPI) a high-intensity UCN source is projected at the WWR-M reactor, which must provide UCN density 2-3 orders of magnitude higher than existing sources. The results of simulations of the designed experimental scheme show that the sensitivity can be increased by ∼ 10–40 times compared to sensitivity of previous experiment depending on the model of neutron reflection from walls.

  13. 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

  14. 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.

  15. 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…

  16. 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…

  17. 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.

  18. Boiling water scarification plus stratification improves germination of Iliamna rivularis (Malvaceae) seeds

    Treesearch

    Katri Himanen; Markku Nygren; R. Kasten Dumroese

    2012-01-01

    Scarification with boiling water plus stratification was most effective in improving germination of Iliamna rivularis (Douglas ex Hook.) Greene (Malvaceae) in an experiment that compared 3 treatments. Seeds from 15 sites representing 5 western US states were used in the experiment. Initial response of the seedlots to the treatments was similar, apart from one seedlot....

  19. Multiphysics modeling of two-phase film boiling within porous corrosion deposits

    SciTech Connect

    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. - Highlights: • A two-phase model of CRUD's effects on fuel cladding is developed and improved. • This model eliminates the formerly erroneous assumption of wick boiling. • Higher fuel cladding temperatures are predicted when accounting for two-phase flow. • Double-peaks in thermal conductivity vs. heat flux in experiments are explained. • A “double dryout” mechanism in CRUD is proposed based on the model and experiments.

  20. Static-flow-instability in subcooled flow boiling in wide rectangular parallel channels

    SciTech Connect

    Siman-Tov, M.; Felde, D.K.; McDuffee, J.L.; Yoder, G.L. Jr.

    1995-12-31

    The Advanced Neutron Source (ANS) is a state-of-the-art research reactor facility that will be built at the Oak Ridge National Laboratory (ORNL) and is designed to become the world`s most advanced thermal neutron flux source for scientific experiments. Therefore, the core of the ANS reactor (ANSR) must be designed to accommodate very high power densities using very high coolant mass fluxes and subcooling levels, The nominal average and peak heat fluxes in the ANSR are approximately 6 and 12 MW/M{sup 2}, respectively, with a nominal total thermal power of 303 MW. Highly subcooled heavy-water coolant (1.7 MPa and 85{degrees}C at the core exit) flows vertically upward at a very high mass flux of almost 27 Mg/M{sup 2}-s. The cooling channels in each fuel assembly are all parallel and share common inlet and outlet plenums, effectively imposing a common pressure drop across all the channels. This core configuration is subject to flow excursion (FE) and/or flow instability that may occur once boiling is initiated in any one of the channels. The FE phenomenon constitutes a different thermal limit than a true critical heat flux (CHF) or departure from nucleate boiling (DNB). In such a system, initiation of boiling in one of the channels (i.e., the hot channel) can result in flow redistribution to the other cooler channels. This report details testing to document this phenomenon.

  1. Preliminary results of the US pool-boiling coils from the IFSMTF full-array tests

    SciTech Connect

    Lue, J.W.; Dresner, L.; Lubell, M.S.; Luton, J.N.; McManamy, T.J.; Shen, S.S.

    1986-01-01

    The Large Coil Task to develop superconducting magnets for fusion reactors, is now in the midst of full-array tests in the International Fusion Superconducting Magnet Test Facility at Oak Ridge National Laboratory. Included in the test array are two pool-boiling coils designed and fabricated by US manufacturers, General Dynamics/Convair Division and General Electric/Union Carbide Corporation. So far, both coils have been energized to full design currents in the single-coil tests, and the General Dynamics coil has reached the design point in the first Standard-I full-array test. Both coils performed well in the charging experiments. Extensive heating tests and the heavy instrumentation of these coils have, however, revealed some generic limitations of large pool-boiling superconducting coils. Details of these results and their analyses are reported.

  2. 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

  3. 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.

  4. 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

  5. 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.

  6. 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

  7. High flux film and transition boiling

    SciTech Connect

    Witte, L.C.

    1993-02-01

    An investigation was conducted on the potential for altering the boiling curve through effects of high velocity and high subcooling. Experiments using water and Freon-113 flowing over cylindrical electrical heaters in crossflow were made to see how velocity and subcooling affect the boiling curve, especially the film and transition boiling regions. We sought subcooling levels down to near the freezing points of these two liquids to prove the concept that the critical heat flux and the minimum heat flux could be brought together, thereby averting the transition region altogether. Another emphasis was to gain insight into how the various boiling regions could be represented mathematically on various parts of heating surface. Motivation for the research grew out of a realization that the effects of very high subcooling and velocity might be to avert the transition boiling altogether so that the unstable part of the boiling curve would not limit the application of high flux devices to temperatures less than the burnout temperatures. Summaries of results from the study are described. It shows that the potential for averting, the transition region is good, and points the way to further research that is needed to demonstrate the potential.

  8. Numerical simulation of pool boiling for steady state and transient heating

    SciTech Connect

    He, Ying; Shoji, Masahiro; Maruyama, Shigeo

    1999-07-01

    It's believed that the macrolayer plays an important role in nucleate and transition boiling heat transfer at high heat flux. Many experiments have been carried out to support the macrolayer evaporation model, however, little has been conducted in the numerical simulation of boiling heat transfer. In this study, based on the macrolayer evaporation model of Maruyama et al. (1992), a numerical simulation of pool boiling for steady state was carried out. The key points of the simulation are: (1) It is modeled that the macrolayer containing vapor stems occupies the region immediately next to the wall and that the vapor stems are formed on the active cavity sites. (2) Not only does the evaporation occur at the vapor bubble-macrolayer interface, but also at the liquid-vapor stem interface. (3) The macrolayers form periodically. No liquid is supplied to the macrolayers during the hovering period. While the vapor mass departs from the surface, the macrolayers replenish immediately despite the complicity of the transition period between the departures of two vapor masses. The major results are: (1) The boiling curves of water and FC-72 (C{sub 6}F{sub 14}) were reasonably predicted. (2) The temporal variations in surface temperature for different boiling regimes were obtained. Secondly, the simulation of transient pool boiling was conducted. It was realized with following assumptions: (1) The macrolayer evaporation model can be extended to the transient pool boiling. The macrolayer forms cyclically and its thickness is determined by the surface heat flux when the vapor mass takes off. (2) One-dimensional transient heat conduction within the heater coupled with the macrolayer model was considered. Being employed explicit FDM, the instantaneous surface temperature can be obtained. Therefore, the instantaneous heat flux can be calculated by applying the surface temperature into the macrolayer model. (3) In the transition-boiling regime, the initial thickness of macrolayer was

  9. Peptide derived from desalinated boiled tuna extract inhibits adipogenesis through the downregulation of C/EBP-α and PPAR-γ in 3T3-L1 adipocytes.

    PubMed

    Kim, Young-Min; Kim, Eun-Young; Kim, In-Hye; Nam, Taek-Jeong

    2015-05-01

    Recently, obesity has increased due to a variety of reasons, including the availability of 'fast food' and high-fat diets. Developing anti-obesity functional drugs and foods from natural sources may offer solutions to this global concern. Generally, tuna is a high-protein, low-fat and low-calorie food with various bioactive effects. It may improve memory, reduce cholesterol levels and positively affect the development of brain cells. In this study, we screened the anti-obesity potential of peptides derived from tuna protein. We then observed protein bands by the Coomassie blue staining of a sodium dodecyl sulfate-polyacrylamide gel electrophoresis (SDS-PAGE) gel. The protein mixture was concentrated and desalted using in-gel trypsin digestion and a C18 nano column and Poros R2 reversed-phase preparation, prior to quadrupole time-of-flight mass spectrometry (Q-TOF MS/MS). We screened the peptides for their ability to affect adipogenesis in 3T3-L1 adipocytes. We also measured glucose uptake, triglyceride levels and lipid droplets using Oil Red O staining. As a result, we confirmed that one peptide inhibited adipocyte differentiation. We also observed the expression of obesity-related genes by western blot analysis and reverse transcription-polymerase chain reaction. The peptide from the tuna extract significantly reduced the expression levels of CCAAT/enhancer-binding protein α (C/EBP-α) and peroxisome proliferator-activated receptor-γ (PPAR-γ) adipocyte marker genes. Thus, our data suggest that this peptide from boiled tuna extract reduces lipid components and adipogenesis in 3T3-L1 cells, and these characteristics may be of value in the development of anti-obesity foods.

  10. 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.

  11. Boiling water reactor uranium utilization improvement potential

    SciTech Connect

    Wei, P.; Crowther, R.L.; Fennern, L.E.; Savoia, P.J.; Specker, S.R.; Tilley, R.M.; Townsend, D.B.; Wolters, R.A.

    1980-06-01

    This report documents the results of design and operational simulation studies to assess the potential for reduction of BWR uranium requirements. The impact of the improvements on separative work requirements and other fuel cycle requirements also were evaluated. The emphasis was on analysis of the improvement potential for once-through cycles, although plutonium recycle also was evaluated. The improvement potential was analyzed for several design alternatives including axial and radial natural uranium blankets, low-leakage refueling patterns, initial core enrichment distribution optimization, reinsert of initial core discharge fuel, preplanned end-of-cycle power coastdown and feedwater temperature reduction, increased discharge burnup, high enrichment discharge fuel rod reassembly and reinsert, lattice and fuel bundle design optimization, coolant density spectral shift with flow control, reduced burnable absorber residual, boric acid for cold shutdown, six-month subcycle refueling, and applications of a once-through thorium cycle design and plutonium recycle.

  12. 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.

  13. 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.

  14. Resolving octant degeneracy at LBL experiment by combining Daya Bay reactor setup

    NASA Astrophysics Data System (ADS)

    Bora, Kalpana; Dutta, Debajyoti

    2014-03-01

    Long baseline Experiment(LBL) have promised to be a very powerful experimental setup to study various issues related to Neutrinos. Some ongoing and planned LBL and medium baseline experiments are- T2K, MINOS, NOvA, LBNE, LBNO etc. But, the long baseline experiments are crippled due to presence of some parameter degeneracies, like the Octant -degeneracy. In this work, we first show the presence of Octant degeneracy in LBL experiments and then combine it with Daya Bay Reactor experiment at different values of CP violation phase. We show that the Octant degeneracy in LBNE can be resolved completely with this proposal.

  15. Computer network that assists in the planning, execution and evaluation of in-reactor experiments

    SciTech Connect

    Bauer, T.H.; Froehle, P.H.; August, C.; Baldwin, R.D.; Johanson, E.W.; Kraimer, M.R.; Simms, R.; Klickman, A.E.

    1985-01-01

    For over 20 years complex, in-reactor experiments have been performed at Argonne National Laboratory (ANL) to investigate the performance of nuclear reactor fuel and to support the development of large computer codes that address questions of reactor safety in full-scale plants. Not only are computer codes an important end-product of the research, but computer analysis is also involved intimately at most stages of experiment planning, data reduction, and evaluation. For instance, many experiments are of sufficiently long duration or, if they are of brief duration, occur in such a purposeful sequence that need for speedy availability of on-line data is paramount. This is made possible most efficiently by computer assisted displays and evaluation. A purposeful linking of main-frame, mini, and micro computers has been effected over the past eight years which greatly enhances the speed with which experimental data are reduced to useful forms and applied to the relevant technological issues. This greater efficiency in data management led also to improvements in the planning and execution of subsequent experiments. Raw data from experiments performed at INEL is stored directly on disk and tape with the aid of minicomputers. Either during or shortly after an experiment, data may be transferred, via a direct link, to the Illinois offices of ANL where the data base is stored on a minicomputer system. This Idaho-to-Illinois link has both enhanced experiment performance and allowed rapid dissemination of results.

  16. Results of a two-position reactor neutrino-oscillation experiment

    SciTech Connect

    Greenwood, Z.D.; Kropp, W.R.; Mandelkern, M.A.; Nakamura, S.; Pasierb-Love, E.L.; Price, L.R.; Reines, F.; Riley, S.P.; Sobel, H.W.; Baumann, N.; Gurr, H.S.

    1996-06-01

    A search for vacuum neutrino oscillations in the range {Delta}{ital m}{sup 2}{approx_gt}0.01 eV{sup 2} with sin{sup 2}2{theta}{approx_gt}0.05 was performed with a detector placed at two different distances (18 and 24 m) from a production reactor at the Savannah River Site. The results observed are inconsistent with oscillations in this region and consistent with other reactor experiments. {copyright} {ital 1996 The American Physical Society.}

  17. 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.

  18. 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.

  19. Improving reactor models with a precision beta spectroscopy experiment of the 235U fission spectrum

    NASA Astrophysics Data System (ADS)

    Boswell, Melissa; Gehring, Amanda; Ullmann, John; Haines, Todd; Devlin, Matthew; Elliott, Steven; Rielage, Keith; Goett, John; White, Brandon

    2016-09-01

    The reactor neutrino anomaly could be an indication of non-standard neutrino physics models (e.g. sterile neutrinos), or the discrepancy could be the result of uncertainties in the modeled reactor neutrino spectrum. Understanding the aggregate β spectrum is extremely important for improving the confidence in the underlying predicted reactor neutrino spectrum. The original β spectrum measurement was preformed at the Institut Laue-Langevin (ILL) reactor in the early 1980's and suffered from a number of limitations. A recent attempt at reproducing the ILL experiment concluded that their uncertainties were driven by a low signal to noise ratio, and that the best way to decrease the uncertainties was to conduct the experiment at a neutron beam line similar to the one at Los Alamos. In this talk I will discuss the advantages of reproducing this experiment at the Lujan center at LANL, our proposed experimental setup, and finally steps we are taking to evaluate the systematics associated with these measurements at an accelerator-based neutron beam. Furthermore we will also discuss additional measurement with 238U and 239Pu that are also important to the non-proliferation communities.

  20. 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.

  1. Background study for the KamLAND reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    Ichimura, K.; Minekawa, Y.

    2008-07-01

    One of the goals of the KamLAND experiment is a search for anti-neutrino oscillation via inverse β decay with the characteristic delayed-coincidence method in the liquid scintillator. For a more precise measurement than previous KamLAND result [1], we have improved the background estimations of (α, n) and fast neutrons. We present the estimated number of backgrounds in our data set from Mar. 2002 to May 2007.

  2. 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.

  3. 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)

  4. 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.

  5. 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

  6. 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)

  7. 78 FR 37595 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-21

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR.... Kathy Weaver, Acting Chief, Technical Support Branch, Advisory Committee on Reactor Safeguards....

  8. 76 FR 5218 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor... inconvenience. Dated: January 24, 2011. Antonio Dias, Chief, Reactor Safety Branch B, Advisory Committee...

  9. 77 FR 59678 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-28

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Branch, Advisory Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  10. 78 FR 20959 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-08

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting.... Antonio Dias, Technical Advisor, Advisory Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  11. Pile noise experiment in MINERVE reactor to estimate kinetic parameters using various data processing methods

    SciTech Connect

    Geslot, Benoit; Gruel, Adrien; Pepino, Alexandra; Di Salvo, Jacques; Izarra, Gregoire de; Jammes, Christian; Destouches, Christophe; Blaise, Patrick

    2015-07-01

    MINERVE is a two-zone pool type zero power reactor operated by CEA (Cadarache, France). Kinetic parameters of the core (prompt neutron decay constant, delayed neutron fraction, generation time) have been recently measured using various pile noise experimental techniques, namely Feynman-α, Rossi-α and Cohn-α. Results are discussed and compared to each other's. The measurement campaign has been conducted in the framework of a tri-partite collaboration between CEA, SCK.CEN and PSI. Results presented in this paper were obtained thanks to a time-stamping acquisition system developed by CEA. PSI performed simultaneous measurements which are presented in a companion paper. Signals come from two high efficiency fission chambers located in the graphite reflector next to the core driver zone. Experiments were conducted at critical state with a reactor power of 0.2 W. The core integral fission rate is obtained from a calibrated miniature fission chamber located at the center of the core. Other results obtained in two sub-critical configurations will be presented elsewhere. Best estimate delayed neutron fraction comes from the Cohn-α method: 747 ± 15 pcm (1σ). In this case, the prompt decay constant is 79 ± 0.5 s{sup -1} and the generation time is 94.5 ± 0.7 μs. Other methods give consistent results within the confidence intervals. Experimental results are compared to calculated values obtained from a full 3D core modeling with the CEA-developed Monte Carlo code TRIPOLI4.9 associated with its continuous energy JEFF3.1.1-based library. A very good agreement is observed for the calculated delayed neutron fraction (748.7 ± 0.4 pcm at 1σ), that is a difference of -0.3% with the experiment. On the contrary, a 10% discrepancy is observed for the calculated generation time (104.4 ± 0.1 μs at 1σ). (authors)

  12. 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.

  13. 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.

  14. 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.

  15. Numerical simulation of pool boiling of a Lennard-Jones liquid

    NASA Astrophysics Data System (ADS)

    Inaoka, Hajime; Ito, Nobuyasu

    2013-09-01

    We performed a numerical simulation of pool boiling by a molecular dynamics model. In the simulation, a liquid composed of Lennard-Jones particles in a uniform gravitational field is heated by a heat source at the bottom of the system. The model successfully reproduces the change in regimes of boiling from nucleate boiling to film boiling with the increase of the heat source temperature. We present the pool boiling curve by the model, whose general behavior is consistent with those observed in experiments of pool boiling.

  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. 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.

  18. Reactor for exothermic reactions

    DOEpatents

    Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

    1993-03-02

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  19. Reactor for exothermic reactions

    DOEpatents

    Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

    1993-01-01

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  20. Numerical Investigation of Boiling

    NASA Astrophysics Data System (ADS)

    Sagan, Michael; Tanguy, Sebastien; Colin, Catherine

    2012-11-01

    In this work, boiling is numerically investigated, using two phase flow direct numerical simulation based on a level set / Ghost Fluid method. Nucleate boiling implies both thermal issue and multiphase dynamics issues at different scales and at different stages of bubble growth. As a result, the different phenomena are investigated separately, considering their nature and the scale at which they occur. First, boiling of a static bubble immersed in an overheated liquid is analysed. Numerical simulations have been performed at different Jakob numbers in the case of strong density discontinuity through the interface. The results show a good agreement on bubble radius evolution between the theoretical evolution and numerical simulation. After the validation of the code for the Scriven test case, interaction of a bubble with a wall is studied. A numerical method taking into account contact angle is evaluated by comparing simulations of the spreading of a liquid droplet impacting on a plate, with experimental data. Then the heat transfer near the contact line is investigated, and simulations of nucleate boiling are performed considering different contact angles values. Finally, the relevance of including a model to take into account the evaporation of the micro layer is discussed.