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Sample records for boiling reactor experiment 4

  1. Mark I 1/5-scale boiling water reactor pressure suppression experiment. Quick-look report for test numbers 1. 0(a) and 1. 0(b) performed on March 4 and 8, 1977

    Energy Technology Data Exchange (ETDEWEB)

    McCauley, E.W.; Pitts, J.H.

    1977-03-16

    The experimental results obtained from pressure suppression experiment numbers 1.0(a) and 1.0(b) that were performed on the Lawrence Livermore Laboratory's /sup 1///sub 5/-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility are summarized.

  2. Gamma heated subassembly for sodium boiling experiments

    Energy Technology Data Exchange (ETDEWEB)

    Artus, S.C.

    1975-01-01

    The design of a system to boil sodium in an LMFBR is examined. This design should be regarded as a first step in a series of boiling experiments. The reactor chosen for the design of the boiling apparatus is the Experimental Breeder Reactor-II (EBR-II), located at the National Reactor Testing Station in Idaho. Criteria broadly classified as design objectives and design requirements are discussed.

  3. Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

    Energy Technology Data Exchange (ETDEWEB)

    Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

    1977-10-11

    An accurate Mark I /sup 1///sub 5/-scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Liao, Huafei [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-12-01

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

  5. Boiling water reactors with Uranium-Plutonium mixed oxide fuel. Report 1: Accuracy of the nuclide concentrations calculated by CASMO-4

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1999-07-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. These CMS (Core Management System) programs have been extensively compared with both measurements and reference codes. Nevertheless some data are proprietary in particular the comparison of the calculated nuclide concentrations versus experiments (because of the cost of this kind of experimental study). This is why this report describes such a comparative investigation carried out with a General Electric 7x7 BWR bundle. Unfortunately, since some core history parameters were unknown, a lot of hypotheses have been adopted. This invokes sometimes a significant discrepancy in the results without being able to determine the origin of the differences between calculations and experiments. Yet one can assess that, except for four nuclides - Plutonium-238, Curium-243, Curium-244 and Cesium-135 - for which the approximate power history (history effect) can be invoked, the accuracy of the calculated nuclide concentrations is rather good if one takes the numerous approximations into account.

  6. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

    A methodology is presented to evaluate the stability properties of Boiling Water Reactors based on a reduced order model, power measurements, and a non-linear estimation technique. For a Boiling Water Reactor, the feedback reactivity imposed by the thermal-hydraulics has an important effect in the system stability, where the dominant contribution to this feedback reactivity is provided by the void reactivity. The feedback reactivity is a function of the operating conditions of the system, and cannot be directly measured. However, power measurements are relatively easy to obtain from the nuclear instrumentation and process computer, and are used in conjunction with a reduced order model to estimate the gain of the thermal-hydraulics feedback using an Extended Kalman Filter. The reduced order model is obtained by estimating the thermal-hydraulic transfer function from the frequency-domain BWR code LAPUR, and the stability properties are evaluated based on the pair of complex conjugate eigenvalues. Because of the recursive nature of the Kalman Filter, an estimate of the decay ratio is generated every sampling time, allowing continuous estimation of the stability parameters. A test platform based on a nuclear-coupled boiling channel is developed to validate the capability of the BWR stability monitoring methodology. The thermal-hydraulics for the boiling channel is modeled and coupled with neutron kinetics to analyze the non-linear dynamics of the closed-loop system. The model uses point kinetics to study core-wide oscillations, and normalized modal kinetics are introduced to study out-of-phase oscillations. The coolant flow dynamics is dominant in the power fluctuations observed by in-core nuclear instrumentation, and additive white noise is added to the solution for the channel flow in the thermal-hydraulic model to generate noisy power time series. The operating conditions of the channel can be modified to accommodate a wide range of stability conditions

  7. Development of neutron radiography facility for boiling two-phase flow experiment in Kyoto University Research Reactor

    Science.gov (United States)

    Saito, Y.; Sekimoto, S.; Hino, M.; Kawabata, Y.

    2011-09-01

    To visualize boiling two-phase flow at high heat flux by using neutron radiography, a new neutron radiography facility was developed in the B-4 beam hole of KUR. The B-4 beam hole is equipped with a supermirror neutron guide tube with a characteristic wavelength of 1.2 Å, whose geometrical parameters of the guide tube are: 11.7 m total length and 10 mm wide ×74 mm high beam cross-section. The total neutron flux obtained from the KUR supermirror guide tube is about 5×10 7 n/cm 2 s with a nominal thermal output of 5 MW of KUR, which is about 100 times what is obtainable with the conventional KUR neutron radiography facility (E-2 beam hole). In this study a new imaging device, an electric power supply (1200 A, 20 V), and a thermal hydraulic loop were installed. The neutron source, the beam tube, and the radiography rooms are described in detail and the preliminary images obtained at the developed facility are shown.

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

    Science.gov (United States)

    Seiler, J. M.; Rameau, B.

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

  9. Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

    Energy Technology Data Exchange (ETDEWEB)

    Rice, R.E.

    1976-09-01

    Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

  10. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

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

  11. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

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

  12. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    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.

  14. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition, numerica

  15. Self-Sustaining Thorium Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  16. Electrically Driven Liquid Film Boiling Experiment

    Science.gov (United States)

    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.

  17. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    OpenAIRE

    Matjaž Leskovar; Mitja Uršič

    2016-01-01

    A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In ...

  18. Bubble Behavior in Nucleate Boiling Experiment Aboard the Space Shuttle

    OpenAIRE

    Koeln, Justin P.; Boulware, Jeffrey C.; Ban, Heng

    2009-01-01

    Boiling dynamics in microgravity need to be better understood before heat transfer systems based on boiling mechanism can be developed for space applications. This paper presents the results of a nucleate boiling experiment aboard Space Shuttle Endeavor (STS- 108). The experiment utilized nickel-chromium resistance wire to boil water in microgravity, and the data was recorded with a CCD camera and six thermistors. This data was analyzed to determine the behavior of bubble formation, detachmen...

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

    Science.gov (United States)

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

    2016-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

    2016-09-11

    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: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. 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.

  1. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security Requirements 1.0 Background The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by the Dairyland Power Cooperative (DPC). The LACBWR was a nuclear power plant of nominal 50 Mw electrical...

  2. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI AGENCY...) requesting exemptions from certain security requirements in Title 10 of the Code Federal Regulations (10 CFR) 73.55, for the LaCrosse Boiling Water Reactor (LACBWR). This Environmental Assessment (EA) has...

  3. Camera Inspection Arm for Boiling Water Reactors - 13330

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Scott; Rood, Marc [S.A. Technology, 3985 S. Lincoln Ave, Loveland, CO 80537 (United States)

    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)

  4. Overview of the Lockheed Martin Compact Fusion Reactor (CFR) T4B Experiment

    Science.gov (United States)

    McGuire, Thomas

    2016-10-01

    The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. The CFR uses a diamagnetic, high beta, magnetically encapsulated, linear ring cusp plasma confinement scheme. The goal of the T4B experiment is to demonstrate a suitable plasma target for heating experiments and to characterize the behavior of plasma sources in the CFR configuration. The design of the T4B experiment will be presented, including discussion of predicted behavior, plasma sources, heating mechanisms, diagnostics suite and relevant numerical modeling. ©2016 Lockheed Martin Corporation. All Rights Reserved.

  5. Construction of the advanced boiling water reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Natsume, Nobuo; Noda, Hiroshi [Tokyo Electric Power Co. (Japan). Nuclear Power Plant Construction Dept.

    1996-07-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-01

    This report documents the results of 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.

  7. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

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

  8. Boiling Experiment Facility for Heat Transfer Studies in Microgravity

    Science.gov (United States)

    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.

  9. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  10. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Andrey [Paul Scherrer Institut, Villigen (Switzerland); Degueldre, Claude, E-mail: claude.degueldre@psi.ch [Paul Scherrer Institut, Villigen (Switzerland); Kaufmann, Wilfried [Kernkraftwerk Leibstadt, Leibstadt (Switzerland)

    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.

  12. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  13. Reactor Neutrino Experiments

    OpenAIRE

    Cao, Jun

    2007-01-01

    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measu...

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

    Energy Technology Data Exchange (ETDEWEB)

    Buckner, M.R.

    2001-06-26

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

  15. Future Reactor Experiments

    OpenAIRE

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measure...

  16. Oxygen suppression in boiling water reactors. Quarterly report 3, April 1-June 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Burley, E.L.

    1978-12-01

    Boiling water reactors (BWR's) generally use high purity, no-additive feedwater. The primary recirculating coolant is neutral pH and contains 100 to 300 ppB oxygen and stoichiometrically related dissolved hydrogen. However, oxygenated water increases austenitic stainless steel susceptibility to intergranular stress-corrosion cracking (IGSCC) when other requisite factors such as stress and sensitization are present. Thus, reduction or elimination of the oxygen in BWR water may preclude cracking incidents. This program is to perform an in-depth engineering evaluation of the potential suppression additives supported by critical experiments where required to resolve substantive uncertainties. On the basis of the engineering evaluation, the optimum oxygen suppression technique will be selected and a specific BWR plant recommended for an extended (3-year) plant demonstration experiment.

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

    Science.gov (United States)

    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.

  18. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security... Dairyland Power Cooperative (DPC). The LACBWR was a nuclear power plant of nominal 50 Mw electrical output... from the regulations in part 73 as it determines are authorized by law and will not endanger life...

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

    Science.gov (United States)

    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

  20. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  1. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gavilian-Moreno, Carlos [Iberdrola Generacion, S.A., Cofrentes Nuclear Power Plant, Project Engineering Department, Paraje le Plano S/N, Valencia (Spain); Espinosa-Paredes, Gilberto [Area de ingeniera en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Mexico city (Mexico)

    2016-04-15

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-12-01

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

  5. Local stability tests in Dresden 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  8. Non normal modal analysis of oscillations in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suarez-Antola, Roberto, E-mail: roberto.suarez@miem.gub.uy [Ministerio de Industria, Energia y Mineria (MIEM), Montevideo (Uruguay); Flores-Godoy, Jose-Job, E-mail: job.flores@ibero.mx [Universidad Iberoamericana (UIA), Mexico, DF (Mexico). Dept. de Fisica Y Matematicas

    2013-07-01

    The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  10. The D&D of the Experimental Boiling Water Reactor (EBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Boling, L.E.; Yule, T.J.; Bhattacharyya, S.K.

    1996-03-01

    Argonne National Laboratory has completed the D&D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D&D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, M.P. [Instituto de Estudos Avancados - CTA, Sao Paolo (Brazil); Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    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.

  12. Boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Tokarev, Yu.I.; Sokolov, I.N.; Skvortsov, S.A.; Sidorov, A.M.; Krauze, L.V.

    1978-04-01

    The possibility of using a boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant (CHPP) was considered, with design features of the reactor intended for a two-purpose plant. A prestressed reinforced concrete vessel and integral arrangement of the primary circuit ensured reliability of the atomic CHPP using various CHPP flowsheets.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-09-01

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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-02

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

  19. Numerical simulation and artificial neural network modeling of natural circulation boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Garg, A. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039, Assam (India); Sastry, P.S. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039, Assam (India); Pandey, M. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039, Assam (India)]. E-mail: manmohan@iitg.ac.in; Dixit, U.S. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039, Assam (India); Gupta, S.K. [Atomic Energy Regulatory Board, Mumbai 400085 (India)

    2007-02-15

    Numerical simulation of natural circulation boiling water reactor is important in order to study its performance for different designs and under various off-design conditions. Numerical simulations can be performed by using thermal-hydraulic codes. Very fast numerical simulations, useful for extensive parametric studies and for solving design optimization problems, can be achieved by using an artificial neural network (ANN) model of the system. In the present work, numerical simulations of natural circulation boiling water reactor have been performed with RELAP5 code for different values of design parameters and operational conditions. Parametric trends observed have been discussed. The data obtained from these simulations have been used to train artificial neural networks, which in turn have been used for further parametric studies and design optimization. The ANN models showed error within {+-}5% for all the simulated data. Two most popular methods, multilayer perceptron (MLP) and radial basis function (RBF) networks, have been used for the training of ANN model. Sequential quadratic programming (SQP) has been used for optimization.

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

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

  1. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2013-12-15

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m{sup 2} s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Joseph, E-mail: joseph.nielsen@inl.gov [Idaho National Laboratory, 1955 N. Fremont Avenue, P.O. Box 1625, Idaho Falls, ID 83402 (United States); University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Tokuhiro, Akira [University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Hiromoto, Robert [University of Idaho, Department of Computer Science, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Tu, Lei [University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States)

    2015-12-15

    Highlights: • Dynamic Event Tree solutions have been optimized using the Branch-and-Bound algorithm. • A 60% efficiency in optimization has been achieved. • Modeling uncertainty within a risk-informed framework is evaluated. - Abstract: 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

  3. Thermionic plasma injection for the Lockheed Martin T4 Compact Fusion Reactor experiment

    Science.gov (United States)

    Heinrich, Jonathon

    2015-11-01

    Lockheed Martin's Compact Fusion Reactor (CFR) concept relies on diamagnetic confinement in a magnetically encapsulated linear ring cusp geometry. Plasma injection into cusp field configurations requires careful deliberation. Previous work has shown that axial injection via a plasma gun is capable of achieving high-beta conditions in cusp configurations. We present a pulsed, high power thermionic plasma source and the associated magnetic field topology for plasma injection into the caulked-cusp magnetic field. The resulting plasma fueling and cross-field diffusion is discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  6. Fuel lattice design in a boiling water reactor using an ant-colony-based system

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis, E-mail: joseluis.montes@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Facultad de Ciencias, Universidad Autonoma del Estado de Mexico (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Ortiz, Juan Jose, E-mail: juanjose.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Perusquia, Raul, E-mail: raul.perusquia@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico)

    2011-06-15

    Research highlights: > We present an ant-colony-based system for BWR fuel lattice design and optimization. > Assessment of candidate solutions at 0.0 MWd/kg {sup 235}U seems to have a limited scope. > Suitable heuristic rules enable more realistic fuel lattice designs. > The election of the objective has a large impact in CPU time. > ACS enables an important decrease of the initial average U-235 enrichment. - Abstract: This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U{sup 235} enrichment and Gd{sub 2}O{sub 3} concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 x 10 BWR fuel lattice with an initial average U{sup 235} enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U{sup 235} enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U{sup 235} enrichment; whereas, the k-infinity was inside the {+-}100 pcm's range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a

  7. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Commission (NRC) is issuing a new regulatory guide (RG), 1.79.1, ``Initial Test Program of Emergency Core... System (ADAMS): You may access publicly available documents online in the NRC Library at...

  8. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Science.gov (United States)

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Analysis of pressure oscillations and safety relief valve vibrations in the main steam system of a Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Galbally, David, E-mail: dgalbally@innomerics.com [Innomerics, Calle San Juan de la Cruz 2, 28223 Madrid (Spain); García, Gonzalo [Alava Ingenieros, Calle Albasanz 16, 28037 Madrid (Spain); Hernando, Jesús; Sánchez, Juan de Dios [Iberdrola, Calle Tomás Redondo 1, 28033 Madrid (Spain); Barral, Marcos [Alava Ingenieros, Calle Albasanz 16, 28037 Madrid (Spain)

    2015-11-15

    Highlights: • We analyze the vibratory response of safety relief valves in the main steam system of a Boiling Water Reactor. • We show that valve internals experience acceleration spikes of more than 20 g. • Spikes are caused by impacts between the valve disc and the seating surface of the valve nozzle. • Resonances occur at higher Strouhal numbers than those reported in the literature for tandem side branches. • Valves experience high vibration levels even for resonances caused by second order hydrodynamic modes. - Abstract: Steam flow inside the main steam lines of a Boiling Water Reactor can generate high-amplitude pressure oscillations due to coupling between the separated shear layer at the mouth of the safety relief valves (SRVs) and the acoustic modes of the side branches where the SRVs are mounted. It is known that certain combinations of flow velocities and main steam line geometries are capable of generating self-excited pressure oscillations with very high amplitudes, which can endanger the structural integrity of main steam system components, such as safety valves, or reactor internals such as steam dryers. However, main steam systems may also experience lower amplitude pressure oscillations due, for example, to coupling of higher order hydrodynamic modes with acoustic cavity modes, or to incipient resonances where the free stream velocity is slightly lower than the critical flow velocity required to develop a stable locked-on acoustic resonance. The amplitude of these pressure oscillations is typically insufficient to cause readily observable structural damage to main steam system components, but may still have subtle effects on safety relief valves. The investigation presented in this article focuses on the characterization of the response of SRVs under the effects of pressure oscillations associated with acoustic excitations that are insufficient to cause structural damage to the valves or associated equipment. It is shown that valve

  11. Neutrino Experiments at Reactors

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    Thomas, Justin W.

    2006-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

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

  14. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    Energy Technology Data Exchange (ETDEWEB)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment.

  15. Reactor antineutrino experiments

    OpenAIRE

    Lu, Haoqi

    2014-01-01

    Neutrinos are elementary particles in the standard model of particle physics. There are 3 flavors of neutrinos that oscillate among themselves. Their oscillation can be described by a 3$\\times$3 unitary matrix, containing three mixing angles $\\theta_{12}$, $\\theta_{23}$, $\\theta_{13}$, and one CP phase. Both $\\theta_{12}$ and $\\theta_{23}$ are known from previous experiments. $\\theta_{13}$ was unknown just two years ago. The Daya Bay experiment gave the first definitive non-zero value in 2012...

  16. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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

  20. Improved neutron kinetics for coupled three-dimensional boiling water reactor analysis

    Science.gov (United States)

    Akdeniz, Bedirhan

    The need for a more accurate method of modelling cross section variations for off-nominal core conditions is becoming an important issue with the increased use of coupled three-dimensional (3-D) thermal-hydraulics/neutronics simulations. In traditional reactor core analysis, thermal reactor core calculations are customarily performed with 3-D two-group nodal diffusion methods. Steady-state multi-group transport theory calculations on heterogeneous single assembly domains subject to reflective boundary conditions are normally used to prepare the equivalent two-group spatially homogenized nodal parameters. For steady-state applications, the equivalent nodal parameters are theoretically well-defined; but, for transient applications, the definition of the nodal kinetics parameters, in particular, delayed neutron precursor data is somewhat unclear. The fact that delayed neutrons are emitted at considerably lower energies than prompt neutrons and that this difference cannot be accounted for in a two-group representation is of particular concern. To compensate for this inherent deficiency of the two-group model a correction is applied to the nodal values of the delayed neutron fractions; however, the adequacy of this correction has never been tested thoroughly for Boiling Water Reactor (BWR) applications, especially where the instantaneous thermal-hydraulic conditions play an important role on the core neutron kinetics calculations. This thesis proposes a systematic approach to improve the 3-D neutron kinetics modelling in coupled BWR transient calculations by developing, implementing and validating methods for consistent generation of neutron kinetics and delayed neutron data for such coupled thermal-hydraulics/neutronics simulations.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    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.

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

    Science.gov (United States)

    Karve, Atul A.

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

  4. The antineutrino energy structure in reactor experiments

    CERN Document Server

    Novella, P

    2015-01-01

    The recent observation of an energy structure in the reactor antineutrino spectrum is reviewed. The reactor experiments Daya Bay, Double Chooz and RENO have reported a consistent excess of antineutrinos deviating from the flux predictions, with a local significance of about 4$\\sigma$ between 4 and 6 MeV of the positron energy spectrum. The possible causes of the structure are analyzed in this work, along with the different experimental approaches developed to identify its origin. Considering the available data and results from the three experiments, the most likely explanation concerns the reactor flux predictions and the associated uncertainties. Therefore, the different current models are described and compared. The possible sources of incompleteness or inaccuracy of such models are discussed, as well as the experimental data required to improve their precision.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Automatic boiling water reactor loading pattern design using ant colony optimization algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Wang, C.-D. [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2 Kuang Fu Road, Hsinchu 30013, Taiwan (China); Nuclear Engineering Division, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)], E-mail: jdwang@iner.gov.tw; Lin Chaung [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2 Kuang Fu Road, Hsinchu 30013, Taiwan (China)

    2009-08-15

    An automatic boiling water reactor (BWR) loading pattern (LP) design methodology was developed using the rank-based ant system (RAS), which is a variant of the ant colony optimization (ACO) algorithm. To reduce design complexity, only the fuel assemblies (FAs) of one eight-core positions were determined using the RAS algorithm, and then the corresponding FAs were loaded into the other parts of the core. Heuristic information was adopted to exclude the selection of the inappropriate FAs which will reduce search space, and thus, the computation time. When the LP was determined, Haling cycle length, beginning of cycle (BOC) shutdown margin (SDM), and Haling end of cycle (EOC) maximum fraction of limit for critical power ratio (MFLCPR) were calculated using SIMULATE-3 code, which were used to evaluate the LP for updating pheromone of RAS. The developed design methodology was demonstrated using FAs of a reference cycle of the BWR6 nuclear power plant. The results show that, the designed LP can be obtained within reasonable computation time, and has a longer cycle length than that of the original design.

  8. Remarks on boiling water reactor stability analysis. Pt. 1. Theory and application of bifurcation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lange, Carsten; Hurtado, Antonio [Technische Univ. Dresden (Germany). Chair of Hydrogen and Nuclear Energy; Schuster, Roland [Kernkraftwerk Brunsbuettel GmbH und Co. oHG, Brunsbuettel (Germany); Lukas, Bernard [EnBW Kernkraft GmbH, Philippsburg (Germany). Kernkraftwerk Philippsburg; Aguirre, Carlos [Kernkraftwerk Leibstadt AG, Aargau (Switzerland); Hennig, Dieter

    2012-11-15

    Modern theoretical methods for analysing the stability behaviour of Boiling Water Reactors (BWRs) are relatively reliable. The analysis is performed by comprehensive validated system codes comprising 3D core models and one-dimensional thermal-hydraulic parallel channel models in the frequency (linearized models) or time domain. Nevertheless the spontaneous emergence of stable or unstable periodic orbits as solutions of the coupled nonlinear differential equations determining the stability properties of the coupled thermal-hydraulic and neutron kinetic (highly) nonlinear BWR system is a surprising phenomenon, and it is worth thinking about the mathematical background controlling such behaviour. In particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points, are states of stability, not all nuclear engineers are familiar with. Hence the part I of this paper is devoted to the mathematical background of linear and nonlinear stability analysis and introduces a novel efficient approach to treat the nonlinear BWR stability behaviour with both system codes and so-called (advanced) reduced order models (ROMs). The efficiency of this approach, called the RAM-ROM method, will be demonstrated by some results of stability analyses for different power plants. (orig.)

  9. Recirculation pump discharge line break tests at ROSA-III for a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, M.; Anoda, Y.; Kumamaru, H.; Nakamura, H.; Shiba, M.; Tasaka, K.

    1985-08-01

    Three loss-of-coolant accident (LOCA) tests were conducted at the Rig of Safety Assessment (ROSA)-III test facility, which simulates boiling water reactor (BWR)/6-251 with a volumetric scaling factor of 1/424. The fundamental features of the recirculation pump discharge line break LOCA and the effects of break areas on the features are investigated. It has been confirmed experimentally that the LOCA phenomena in the discharge line break are analogous to those in the suction line break with the same effective choking flow area, which is a sum of the least choking flow areas along the break flow paths and controls the system pressure responses. In general, the maximum effective choking flow area is (A /SUB j/ + A /SUB p/ ) for discharge line breaks and (A /SUB j/ + A /SUB o/ ) for suction line breaks, where A /SUB j/ , A /SUB p/ , and A /SUB o/ are the flow areas of the jet pump drive nozzles, the main recirculation pump discharge nozzle, and the break, respectively. The similarity between the ROSA-III test and a BWR LOCA has been confirmed in the key phenomena by the analyses using the RELAP5/MOD1 code. An atypical behavior is observed in the fuel rod surface temperature transient in the early phase of blowdown due to the limitation of the ROSA-III initial core power.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  11. Bubble spreading during the boiling crisis: modelling and experimenting in microgravity

    Science.gov (United States)

    Nikolayev, V.; Beysens, D.; Garrabos, Y.; Lecoutre, C.; Chatain, D.

    2006-09-01

    Boiling is a very efficient way to transfer heat from a heater to the liquid carrier. We discuss the boiling crisis, a transition between two regimes of boiling: nucleate and film boiling. The boiling crisis results in a sharp decrease in the heat transfer rate, which can cause a major accident in industrial heat exchangers. In this communication, we present a physical model of the boiling crisis based on the vapor recoil effect. Under the action of the vapor recoil the gas bubbles begin to spread over the heater thus forming a germ for the vapor film. The vapor recoil force not only causes its spreading, it also creates a strong adhesion to the heater that prevents the bubble departure, thus favoring the further spreading. Near the liquid-gas critical point, the bubble growth is very slow and allows the kinetics of the bubble spreading to be observed. Since the surface tension is very small in this regime, only microgravity conditions can preserve a convex bubble shape. In the experiments both in the Mir space station and in the magnetic levitation facility, we directly observed an increase of the apparent contact angle and spreading of the dry spot under the bubble. Numerical simulations of the thermally controlled bubble growth show this vapor recoil effect too thus confirming our model of the boiling crisis.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-05-01

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

  15. Bubble spreading during the boiling crisis: modelling and experimenting in microgravity

    CERN Document Server

    Nikolayev, Vadim; Garrabos, Y; Lecoutre, C; Chatain, D

    2016-01-01

    Boiling is a very efficient way to transfer heat from a heater to the liquid carrier. We discuss the boiling crisis, a transition between two regimes of boiling: nucleate and film boiling. The boiling crisis results in a sharp decrease in the heat transfer rate, which can cause a major accident in industrial heat exchangers. In this communication, we present a physical model of the boiling crisis based on the vapor recoil effect. Under the action of the vapor recoil the gas bubbles begin to spread over the heater thus forming a germ for the vapor film. The vapor recoil force not only causes its spreading, it also creates a strong adhesion to the heater that prevents the bubble departure, thus favoring the further spreading. Near the liquid-gas critical point, the bubble growth is very slow and allows the kinetics of the bubble spreading to be observed. Since the surface tension is very small in this regime, only microgravity conditions can preserve a convex bubble shape. In the experiments both in the Mir spa...

  16. Experimental and analytical study of stability characteristics of natural circulation boiling water reactors during startup transient

    Science.gov (United States)

    Woo, Kyoungsuk

    Two-phase natural circulation loops are unstable at low pressure operating conditions. New reactor design relying on natural circulation for both normal and abnormal core cooling is susceptible to different types of flow instabilities. In contrast to forced circulation boiling water reactor (BWR), natural circulation BWR is started up without recirculation pumps. The tall chimney placed on the top of the core makes the system susceptible to flashing during low pressure start-up. In addition, the considerable saturation temperature variation may induce complicated dynamic behavior driven by thermal non-equilibrium between the liquid and steam. The thermal-hydraulic problems in two-phase natural circulation systems at low pressure and low power conditions are investigated through experimental methods. Fuel heat conduction, neutron kinetics, flow kinematics, energetics and dynamics that govern the flow behavior at low pressure, are formulated. A dimensionless analysis is introduced to obtain governing dimensionless groups which are groundwork of the system scaling. Based on the robust scaling method and start-up procedures of a typical natural circulation BWR, the simulation strategies for the transient with and without void reactivity feedback is developed. Three different heat-up rates are applied to the transient simulations to study characteristics of the stability during the start-up. Reducing heat-up rate leads to increase in the period of flashing-induced density wave oscillation and decrease in the system pressurization rate. However, reducing the heat-up rate is unable to completely prevent flashing-induced oscillations. Five characteristic regions of stability are discovered at low pressure conditions. They are stable single-phase, flashing near the separator, intermittent oscillation, sinusoidal oscillation and low subcooling stable regions. Stability maps were acquired for system pressures ranging 100 kPa to 400 kPa. According to experimental investigation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

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

  19. Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, Gilberto, E-mail: gepe@xanum.uam.m [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico D.F., 09340 (Mexico); Verma, Surendra P. [Centro de Investigacion en Energia, Universidad Nacional Autonoma de Mexico, Priv. Xochicalco s/no., Col Centro, Apartado Postal 34, Temixco 62580 (Mexico); Vazquez-Rodriguez, Alejandro [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico D.F., 09340 (Mexico); Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragan 779, Col. Narvarte, Mexico D.F. 03020 (Mexico)

    2010-05-15

    Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.

  20. Oscillation Parameters with forthcoming Reactor Neutrino Experiments

    CERN Document Server

    Lasserre, Thierry

    2010-01-01

    I review the status of the forthcoming reactor neutrino experiments that toe the cutting edge of neutrino oscillation research. Kilometer baseline oscillation experiments (Double Chooz, Daya Bay, and Reno) will soon play a relevant role providing clean information on the last undetermined neutrino mixing angle !13. A 50-70 km baseline reactor neutrino experiment could later provide the best sensitivity to the !12 mixing angle.

  1. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  2. Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deokule, A.P., E-mail: abhijit.deokule1986@gmail.com [Homi Bhabha National Institute, Trombay 400 085, Mumbai (India); Vishnoi, A.K.; Dasgupta, A.; Umasankari, K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Research Centre, Trombay 400 085, Mumbai (India)

    2014-09-15

    Highlights: • Development of 19 pin annular fuel rod cluster. • Reactor physics study of designed annular fuel rod cluster. • Thermal hydraulic study of annular fuel rod cluster. - Abstract: An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

  3. Azcaxalli: A system based on Ant Colony Optimization algorithms, applied to fuel reloads design in a Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel-Estrada, Jaime, E-mail: jaime.esquivel@fi.uaemex.m [Facultad de Ingenieria, Universidad Autonoma del Estado de Mexico, Cerro de Coatepec S/N, Toluca de Lerdo, Estado de Mexico 50000 (Mexico); Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico); Ortiz-Servin, Juan Jose, E-mail: juanjose.ortiz@inin.gob.m [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico); Castillo, Jose Alejandro; Perusquia, Raul [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico)

    2011-01-15

    This paper presents some results of the implementation of several optimization algorithms based on ant colonies, applied to the fuel reload design in a Boiling Water Reactor. The system called Azcaxalli is constructed with the following algorithms: Ant Colony System, Ant System, Best-Worst Ant System and MAX-MIN Ant System. Azcaxalli starts with a random fuel reload. Ants move into reactor core channels according to the State Transition Rule in order to select two fuel assemblies into a 1/8 part of the reactor core and change positions between them. This rule takes into account pheromone trails and acquired knowledge. Acquired knowledge is obtained from load cycle values of fuel assemblies. Azcaxalli claim is to work in order to maximize the cycle length taking into account several safety parameters. Azcaxalli's objective function involves thermal limits at the end of the cycle, cold shutdown margin at the beginning of the cycle and the neutron effective multiplication factor for a given cycle exposure. Those parameters are calculated by CM-PRESTO code. Through the Haling Principle is possible to calculate the end of the cycle. This system was applied to an equilibrium cycle of 18 months of Laguna Verde Nuclear Power Plant in Mexico. The results show that the system obtains fuel reloads with higher cycle lengths than the original fuel reload. Azcaxalli results are compared with genetic algorithms, tabu search and neural networks results.

  4. Verification of the IVA4 film boiling model with the data base of Liu and Theofanous

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1998-01-01

    Part 1 of this work presents a closed analytical solution for mixed-convection film boiling on vertical walls. Heat transfer coefficients predicted by the proposed model and experimental data obtained at the Royal Institute of Technology in Sweden by Okkonen et al are compared. All data predicted are inside the {+-}10% error band, with mean averaged error being below 4% using the slightly modified analytical solution. The solution obtained is recommended for practical applications. The method presented here is used in Part 2 as a guideline for developing model for film boiling on spheres. The new semi-empirical film boiling model for spheres used in IVA4 computer code is compared with the experimental data base obtained by Liu and Theofanous. The data are predicted within {+-}30% error band. (author)

  5. Experimental and numerical stability investigations on natural circulation boiling water reactors

    CERN Document Server

    Marcel, CP

    2007-01-01

    In the design of novel nuclear reactors active systems are replaced by passive ones in order to reduce the risk of failure. For that reason natural circulation is being considered as the primary cooling mechanism in next generation nuclear reactor designs

  6. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  7. Ignition of DME and DME/CH4 at High Pressure: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    The pyrolysis and oxidation of dimethyl ether (DME) and its mixtures with methane were investigated at high pressures (50 and 100 bar) and intermediate temperatures (450–900 K) in a laminar flow reactor. DME pyrolysis started at 825 K (at 50 bar). The onset of DME reaction was detected at 525–550 K...

  8. Ignition of DME and DME/CH4 at High Pressure: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    The pyrolysis and oxidation of dimethyl ether (DME) and its mixtures with methane were investigated at high pressures (50 and 100 bar) and intermediate temperatures (450―900 K) in a laminar flow reactor. DME pyrolysis started at 825 K (at 50 bar). The onset of DME reaction was detected at 525―550 K...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. A Simple Tubular Reactor Experiment.

    Science.gov (United States)

    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)

  11. Double Chooz and Reactor Theta13 Experiments

    CERN Document Server

    ,

    2016-01-01

    This is a contribution paper from the Double Chooz experiment to the special issue of NPB on neutrino oscillations. The physics and history of the reactor theta13 experiments, as well as Double Chooz experiment and its neutrino oscillation analyses are reviewed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cole, C.J.P

    2005-07-01

    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{sup o}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

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

    Science.gov (United States)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Science.gov (United States)

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

    2011-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. The Daya Bay Reactor Neutrino Experiment

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    On Aug.15, 201l, a new large-scale scientific facility in China, Daya Bay Reactor Neutrino Experiment, started to operate. It is located in Daya Bay Nuclear Power Plant in Guangdong Province, around 50kin to both Hong Kong and Shenzhen City. The main scientific goal is to precisely determine the neutrino mixing angle 013 by detecting neutrinos from the reactors at different distances.

  18. The Development of Nufreq-N AN Analytical Model for the Stability Analysis of Nuclear Coupled Density-Wave Oscillations in Boiling Water Nuclear Reactors.

    Science.gov (United States)

    Park, Goon Cherl

    A state-of-the-art one-dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR). The model accounts for phasic slip, distributed spacers, subcooled boiling, space/time -dependent power distributions and distributed heated wall dynamics. In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation. In its final form, this model constitutes a multi -input, multi-output(MIMO) linear system, which features a general nodal neutron kinetics model. Kinetics parameters for use in the kinetics model have been obtained by utilizing self-consistent nodal data and power distributions. The stability characteristics of a typical BWR/4 has been investigated with the Nyquist criterion. The computer implementation of this model, NUFREQ -N, was used for the parametric study of a typical BWR/4 and comparisons were made with existing in-core and out -of-core data. Also, NUFREQ-N was used to analyze the expected stability characteristics of a typical BWR/4. The parametric results revealed important factors influencing BWR stability margin. It was found that NUFREQ -N generally agreed well with out-of-core data. This was especially true for the predicted power-to-flow transfer function, which is the most important transfer function in thermal-hydraulic stability analysis. In the stability analysis of a typical BWR/4 it was found that it is very important to use accurate models of thermal-hydraulic and neutron kinetic phenomena. Moreover, the accuracy of the nuclear input data is extremely important.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Constraining Sterile Neutrinos Using Reactor Neutrino Experiments

    CERN Document Server

    Girardi, Ivan; Ohlsson, Tommy; Zhang, He; Zhou, Shun

    2014-01-01

    Models of neutrino mixing involving one or more sterile neutrinos have resurrected their importance in the light of recent cosmological data. In this case, reactor antineutrino experiments offer an ideal place to look for signatures of sterile neutrinos due to their impact on neutrino flavor transitions. In this work, we show that the high-precision data of the Daya Bay experi\\-ment constrain the 3+1 neutrino scenario imposing upper bounds on the relevant active-sterile mixing angle $\\sin^2 2 \\theta_{14} \\lesssim 0.06$ at 3$\\sigma$ confidence level for the mass-squared difference $\\Delta m^2_{41}$ in the range $(10^{-3},10^{-1}) \\, {\\rm eV^2}$. The latter bound can be improved by six years of running of the JUNO experiment, $\\sin^22\\theta_{14} \\lesssim 0.016$, although in the smaller mass range $ \\Delta m^2_{41} \\in (10^{-4} ,10^{-3}) \\, {\\rm eV}^2$. We have also investigated the impact of sterile neutrinos on precision measurements of the standard neutrino oscillation parameters $\\theta_{13}$ and $\\Delta m^2...

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

    Science.gov (United States)

    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.

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

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  3. New Dosimetric Interpretation of the DV50 Vessel-Steel Experiment Irradiated in the OSIRIS MTR Reactor Using the Monte-Carlo Code TRIPOLI-4®

    Directory of Open Access Journals (Sweden)

    Malouch Fadhel

    2016-01-01

    Full Text Available An irradiation program DV50 was carried out from 2002 to 2006 in the OSIRIS material testing reactor (CEA-Saclay center to assess the pressure vessel steel toughness curve for a fast neutron fluence (E > 1 MeV equivalent to a French 900-MWe PWR lifetime of 50 years. This program allowed the irradiation of 120 specimens out of vessel steel, subdivided in two successive irradiations DV50 n∘1 and DV50 n∘2. To measure the fast neutron fluence (E > 1 MeV received by specimens after each irradiation, sample holders were equipped with activation foils that were withdrawn at the end of irradiation for activity counting and processing. The fast effective cross-sections used in the dosimeter processing were determined with a specific calculation scheme based on the Monte-Carlo code TRIPOLI-3 (and the nuclear data ENDF/B-VI and IRDF-90. In order to put vessel-steel experiments at the same standard, a new dosimetric interpretation of the DV50 experiment has been performed by using the Monte-Carlo code TRIPOLI-4 and more recent nuclear data (JEFF3.1.1 and IRDF-2002. This paper presents a comparison of previous and recent calculations performed for the DV50 vessel-steel experiment to assess the impact on the dosimetric interpretation.

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

    Science.gov (United States)

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

    1988-01-01

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

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

    Science.gov (United States)

    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.

  6. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available 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.

  8. Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Mexico, D.F. (Mexico); Francois, Juan Luis [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico)]. E-mail: jlfl@fi-b.unam.mx; Martin-del-Campo, Cecilia [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana, Avenida San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico)

    2005-04-15

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS 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. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the {sup 233}U build-up. 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 {sup 235}U 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 fuel assembly.

  9. The noncondensable gas effects on loss-of-coolant accident steam condensation loads in boiling water reactor pressure suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Namatame, K.; Shiba, M.; Takeshita, I.

    1983-11-01

    The noncondensable gas effects on the loss-ofcoolant-accident-induced steam condensation loads in the boiling water reactor pressure suppression pool have been investigated with regard to experimental data obtained from a large-scale multivent test program. Previous studies have noted that the presence of the noncondensable gas (air), which initially fills the containment drywell space, stabilizes the direct-contact condensation in the pressure suppression pool and hampers onset of the chugging phenomenon, which induces most significant steam condensation load onto the pool boundary. This was found to be true for the tests with relatively small-break diameters, where the maximum steam mass fluxes in the vent pipe were lower than the upper threshold value for the onset of chugging. However, in the tests with the maximum vent steam mass fluxes moderately higher than the chugging upper threshold value, early depletion of the noncondensable gas tended to result in significant stabilization of steam condensation accompanied by an excursion of temperature of pool water surrounding the vent pipe outlets, which led to a delayed onset of chugging. Due to this combined influence of the noncondensable gas and nonuniform pool temperature, and due to dependence of magnitude of chugging load on the vent steam mass flux, the peak magnitude of the steam condensation load appearing in a blowdown can be very sensitive to the initial and break conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

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

  13. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  14. Sodium Reactor Experiment decommissioning. Final report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Investigation of film boiling thermal hydraulics under FCI conditions. Results of a numerical study

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Dinh, A.T.; Nourgaliev, R.R.; Sehgal, B.R. [Div. of Nuclear Power Safety Royal Inst. of Tech. (RIT), Brinellvaegen 60, 10044 Stockholm (Sweden)

    1998-01-01

    Film boiling on the surface of a high-temperature melt jet or of a melt particle is one of key phenomena governing the physics of fuel-coolant interactions (FCIs) which may occur during the course of a severe accident in a light water reactor (LWR). A number of experimental and analytical studies have been performed, in the past, to address film boiling heat transfer and the accompanying hydrodynamic aspects. Most of the experiments have, however, been performed for temperature and heat flux conditions, which are significantly lower than the prototypic conditions. For ex-vessel FCIs, high liquid subcooling can significantly affect the FCI thermal hydraulics. Presently, there are large uncertainties in predicting natural-convection film boiling of subcooled liquids on high-temperature surfaces. In this paper, research conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning film-boiling thermal hydraulics under FCI condition is presented. Notably, the focus is placed on the effects of (1) water subcooling, (2) high-temperature steam properties, (3) the radiation heat transfer and (4) mixing zone boiling dynamics, on the vapor film characteristics. Numerical investigations are performed using a novel CFD modeling concept named as the local-homogeneous-slip model (LHSM). Results of the analytical and numerical studies are discussed with respect to boiling dynamics under FCI conditions. (author)

  16. ROSA-III double-ended break test series for a loss-of-coolant accident in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tasaka, K.; Anoda, Y.; Koizumi, Y.; Kumamaru, H.; Nakamura, H.; Shiba, M.; Suzuki, M.; Yonomoto, T.

    1985-01-01

    The Rig of Safety Assessment (ROSA) III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure assumption in ECCS is the high-pressure core spray system failure even in a large-break LOCA in a BWR. The measured peak cladding temperature was well below the present safety criterion of 1473 K, even with the single failure assumption in ECCS, and the effectiveness of ECCS for core cooling during a double-ended-break LOCA has been confirmed. The overall agreement between the results calculated by the RELAP4/MOD6/U4/J3 computer code and the experimental results is good. The similarity between the ROSA-III test and a BWR LOCA has been confirmed through the comparison of calculated results for the ROSA-III facility and a BWR system.

  17. Background Studies for the MINER Coherent Neutrino Scattering Reactor Experiment

    CERN Document Server

    Agnolet, G; 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; Ogawa, I; Lang, K; Mahapatra, R; Mandic, V; Martin, R D; Mast, N; McDeavitt, S; Mirabolfathi, N; Mohanty, B; Nakajima, K; Newhouse, J; Newstead, J L; Phan, D; Proga, M; Roberts, A; Rogachev, G; Salazar, R; Sander, J; Senapati, K; Shimada, M; Strigari, L; Tamagawa, Y; Teizer, W; Vermaak, J I C; Villano, A N; Walker, J; Webb, B; Wetzel, Z; Yadavalli, S A

    2016-01-01

    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 meters) 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 to 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Status and Prospects of Reactor Neutrino Experiments

    CERN Document Server

    Kim, Soo-Bong

    2015-01-01

    New generation of three reactor neutrino experiments have made definitive measurements of the smallest neutrino mixing angle theta13 in 2012, based on the disappearance of electron antineutrinos. More precise measurements of the mixing angle have been made as well as the squared mass difference between electron neutrinos. A rather large value of theta13 has opened a new window to find the CP violation phase and to determine the neutrino mass hierarchy. Future reactor experiments, JUNO and RENO50, are proposed to determine the neutrino mass hierarchy and to make highly precise measurements of theta12, the squared mass difference between neutrino masses 2 and 1, and the squared mass difference between electron neutrinos.

  20. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  1. A theoretical model for coupled neutronic-thermohydraulic out-of-phase oscillations in Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bragt, D.D.B. van

    1995-10-01

    A theoretical model for out-of-phase power oscillations in BWRs is proposed. This model describes the dynamic behavior of the neutronic and thermohydraulic subsystems during out-of-phase oscillations, and the coupling of these subsystems via the fuel temperature dynamics and void- and Doppler feedback effects. The zero-power neutron kinetics of the out-of-phase flux density mode is derived by expanding the (time- and space-dependent) neutron flux density in the static solutions of the neutron transport equation. This procedure yields the modal point-kinetic equations for the (first-harmonic) out-of-phase mode. The fuel temperature dynamics is described by a lumped parameter first-order process, characterized by a typical fuel time constant. Using the quasistatic approach, the basic equations of the channel thermohydraulics are derived from the conservation laws of mass and energy and the momentum equation. The momentum equation is coupled with the appropriate boundary condition (constant core pressure drop) for out-of phase oscillations. This procedure yields a set of nonlinear equations describing the dynamic behavior of the boiling boundary, void fraction and mass flux density in the cooling channel. A frequency-domain parametric study confirms that if the out-of-phase mode has a more negative subcriticality, reactor stability increases. On the other hand, a more negative void reactivity coefficient has a destabilizing effect. Besides these two parameters, the fuel time constant was found to be an important parameter determining stability. Where possible, the linearized equations describing the channel thermohydraulics were compare with exact solutions of the governing partial-differential channel equations. This comparison shows that in the frequency range of interest, discrepancies between the proposed quasi-static model and more complicated exact solutions are to be expected. (orig.).

  2. JUNO: A Next Generation Reactor Antineutrino Experiment

    CERN Document Server

    Zhan, Liang

    2015-01-01

    The mass hierarchy and the CP phase are the main focus of the next generation neutrino oscillation experiments. Jiangmen Underground Neutrino Observatory (JUNO), as a medium baseline reactor antineutrino experiment, can determine the neutrino mass hierarchy independent of the CP phase. The physics potential on the mass hierarchy, and other measurements are reviewed. The preliminary design options for a 20~kton detector with an energy resolution of $3\\%/\\sqrt{E_{vis}}$ are illustrated. The main technical challenges on the PMT and scintillator are discussed and the corresponding R\\&D efforts are presented.

  3. Fuel lattice design in a boiling water reactor using a knowledge-based automation system

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2015-11-15

    Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U{sup 235} enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4

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

    Energy Technology Data Exchange (ETDEWEB)

    Palmrose, D.E. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Mandl, R.M. (Siemens AG, Berlin (Germany))

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Neutrino mass hierarchy determination at reactor antineutrino experiments

    CERN Document Server

    Yang, Guang

    2015-01-01

    After the neutrino mixing angle $\\theta_{13}$ has been precisely measured by the reactor antineutrino experiments, one of the most important open questions left in neutrino physics is the neutrino mass hierarchy. Jiangmen Underground Neutrino Observatory (JUNO) is designed to determine the neutrino mass hierarchy (MH) without exploring the matter effect. The JUNO site location is optimized to have the best sensitivity for the mass hierarchy determination. JUNO will employ a 20 kton liquid scintillator detector located in a laboratory 700 meters underground. The excellent energy resolution and PMT coverage will give us an unprecedented opportunity to reach a 3-4 $\\sigma$ precision. In this paper, the JUNO detector design and simulation work will be presented. Also, RENO-50, another medium distance reactor antineutrino experiment, will do a similar measurement. With the efforts of these experiments, it is very likely that the neutrino mass hierarchy will be determined in the next 10 years.

  8. Search for New Physics in reactor and accelerator experiments

    Science.gov (United States)

    Di Iura, A.; Girardi, I.; Meloni, D.

    2016-01-01

    We consider two scenarios of New Physics: the Large Extra Dimensions (LED), where sterile neutrinos can propagate in a (4+d) -dimensional space-time, and the Non Standard Interactions (NSI), where the neutrino interactions with ordinary matter are parametrized at low energy in terms of effective flavour-dependent complex couplings \\varepsilon_{αβ} . We study how these models have an impact on oscillation parameters in reactor and accelerator experiments.

  9. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

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

  12. Investigating the Spectral Anomaly with Different Reactor Antineutrino Experiments

    CERN Document Server

    Buck, Christian; Haser, Julia; Lindner, Manfred

    2015-01-01

    The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in the neutrino flux predictions. 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 flux models and experimental results. We combine experiments at reactors which are highly enriched in ${}^{235}$U with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. 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.

  13. Reactor Simulation for Antineutrino Experiments using DRAGON and MURE

    CERN Document Server

    Jones, C L; Conrad, J M; Djurcic, Z; Fallot, M; Giot, L; Keefer, G; Onillon, A; Winslow, L

    2011-01-01

    Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulations to predict reactor fission rates. Here we present results from the DRAGON and MURE simulation codes and compare them to other industry standards for reactor core modeling. We use published data from the Takahama-3 reactor to evaluate the quality of these simulations against the independently measured fuel isotopic composition. The propagation of the uncertainty in the reactor operating parameters to the resulting antineutrino flux predictions is also discussed.

  14. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1; Dekontamination des RDB inkl. der Einbauten wie Dampftrockner und Wasserabscheider sowie der angeschlossenen Hilfssysteme im deutschen Siedewasserreaktor ISAR 1

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Michael; Sempere Belda, Luis; Basu, Ashim; Topf, Christian [AREVA GmbH, Erlangen (Germany). Abt. Chemistry Services; Erbacher, Thomas; Hiermer, Thomas; Schnurr, Bernhard; Appeldorn, Thomas van [E.ON Kernkraft GmbH, Kernkraftwerk ISAR, Essenbach (Germany). Abt. Maschinentechnik; Volkmann, Christian [ESG Engineering Services GmbH, Greifswald (Germany)

    2015-12-15

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17{sup th}, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  15. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

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

  16. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest National Lab., Richland, WA (United States)

    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.

  18. Uncertainty analysis of fission fraction for reactor antineutrino experiments

    Science.gov (United States)

    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. Determination and Fabrication of New Shield Super Alloys Materials for Nuclear Reactor Safety by Experiments and Cern-Fluka Monte Carlo Simulation Code, Geant4 and WinXCom

    Science.gov (United States)

    Aygun, Bünyamin; Korkut, Turgay; Karabulut, Abdulhalik

    2016-05-01

    Despite the possibility of depletion of fossil fuels increasing energy needs the use of radiation tends to increase. Recently the security-focused debate about planned nuclear power plants still continues. The objective of this thesis is to prevent the radiation spread from nuclear reactors into the environment. In order to do this, we produced higher performanced of new shielding materials which are high radiation holders in reactors operation. Some additives used in new shielding materials; some of iron (Fe), rhenium (Re), nickel (Ni), chromium (Cr), boron (B), copper (Cu), tungsten (W), tantalum (Ta), boron carbide (B4C). The results of this experiments indicated that these materials are good shields against gamma and neutrons. The powder metallurgy technique was used to produce new shielding materials. CERN - FLUKA Geant4 Monte Carlo simulation code and WinXCom were used for determination of the percentages of high temperature resistant and high-level fast neutron and gamma shielding materials participated components. Super alloys was produced and then the experimental fast neutron dose equivalent measurements and gamma radiation absorpsion of the new shielding materials were carried out. The produced products to be used safely reactors not only in nuclear medicine, in the treatment room, for the storage of nuclear waste, nuclear research laboratories, against cosmic radiation in space vehicles and has the qualities.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Wagner, K.C. (Science Applications International Corp., Albuquerque, NM (USA))

    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.

  1. Reactor G1: high power experiments; Experiences a forte puissance

    Energy Technology Data Exchange (ETDEWEB)

    Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Retel, H. [Societe Rateau, D.E.A. (France)

    1957-07-01

    The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

  2. 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

    Science.gov (United States)

    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.

  3. Analysis of flow boiling heat transfer in narrow annular gaps applying the design of experiments method

    Directory of Open Access Journals (Sweden)

    Gunar Boye

    2015-06-01

    Full Text Available The axial heat transfer coefficient during flow boiling of n-hexane was measured using infrared thermography to determine the axial wall temperature in three geometrically similar annular gaps with different widths (s = 1.5 mm, s = 1 mm, s = 0.5 mm. During the design and evaluation process, the methods of statistical experimental design were applied. The following factors/parameters were varied: the heat flux q · = 30 − 190 kW / m 2 , the mass flux m · = 30 − 700 kg / m 2 s , the vapor quality x · = 0 . 2 − 0 . 7 , and the subcooled inlet temperature T U = 20 − 60 K . The test sections with gap widths of s = 1.5 mm and s = 1 mm had very similar heat transfer characteristics. The heat transfer coefficient increases significantly in the range of subcooled boiling, and after reaching a maximum at the transition to the saturated flow boiling, it drops almost monotonically with increasing vapor quality. With a gap width of 0.5 mm, however, the heat transfer coefficient in the range of saturated flow boiling first has a downward trend and then increases at higher vapor qualities. For each test section, two correlations between the heat transfer coefficient and the operating parameters have been created. The comparison also shows a clear trend of an increasing heat transfer coefficient with increasing heat flux for test sections s = 1.5 mm and s = 1.0 mm, but with increasing vapor quality, this trend is reversed for test section 0.5 mm.

  4. Simulation of Reactors for Antineutrino Experiments Using DRAGON

    CERN Document Server

    Winslow, L

    2011-01-01

    From the discovery of the neutrino to the precision neutrino oscillation measurements in KamLAND, nuclear reactors have proven to be an important source of antineutrinos. As their power and our knowledge of neutrino physics has increased, more sensitive measurements have become possible. The next generation of reactor antineutrino experiments require more detailed simulations of the reactor core. Many of the reactor simulation codes are proprietary which makes detailed studies difficult. Here we present the results of the open source DRAGON code and compare it to other industry standards for reactor modeling. We use published data from the Takahama reactor to determine the quality of the simulations. The propagation of the uncertainty to the antineutrino flux is also discussed.

  5. Experimental study of CHF enhancement using Fe{sub 3}O{sub 4} nanofluids in the subcooled boiling region

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young Jae; Kam, Dong Hoon; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    This study may give overall trends of CHF enhancement in the subcooled boiling region. In our experiment, subcooled flow boiling CHF enhancement phenomena in water and nano-coated surface was investigated in mass flux from 1000 to 5000 kg/m{sup 2}s. CHF enhancement of nanoparticles coated tube in DI water increased as exit quality get bigger at same massflux. Various methods to improve CHF characteristics are introduced, especially nanofluids are used for enhancing the CHF. Nanofluids is a colloidal suspension that nanoparticles are mixed with basic fluid. Normally the use of nanofluids as working fluid improves the flow boiling CHF characteristics. Lee et al. already researched the CHF characteristics using nanofluids. As exit quality increased from 0.07 to 0.74, CHF enhancement gradually decreased and approached zero. CHF enhancement was observed when exit quality was low and a DNB-like thermal crisis occurred. But CHF enhancement didn't occur for high exit quality, but LFD-type thermal crisis occurred. Because LFD phenomena are nearly unaffected by the surface conditions, CHF enhancement is not expected for annular flow with high exit quality. Kim et al. performed flow boiling CHF enhancement at subcooled region using alumina-water, zinc-oxide-water and diamond-water nanofluids. The CHF was enhanced by increasing wettability from nanoparticle deposition. CHF enhancement occurred in high mass flux (2000-2500 kg/m{sup 2}s), but CHF enhancement didn't occur in low mass flux (1500 kg/m{sup 2}s). The amount of nanoparticle deposition on each tube can be different during experiments by the several conditions such as deposition time, mass flux and heat flux. So, before the nanofluid experiment conducted, all tube are deposited in same condition of heat flux, concentration and time.

  6. Educational reactor-physics experiments with the critical assemble TCA

    Energy Technology Data Exchange (ETDEWEB)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

  7. High flux film and transition boiling

    Energy Technology Data Exchange (ETDEWEB)

    Witte, L.C.

    1990-01-01

    This report is a bench-scale experiment on transition boiling. The author gives a detailed description on experimental apparatus and conditions. The visual observed boiling phenomena; nucleate boiling and film boiling, and the effect of heat transfer are also elucidated. 10 refs., 11 figs., 1 tab.

  8. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. System Requirements Document for the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Aigner, R.D.

    2000-04-01

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

  10. Reactor Neutrino Experiments with a Large Liquid Scintillator Detector

    CERN Document Server

    Kopp, J F; Merle, A; Rolinec, M

    2007-01-01

    We discuss several new ideas for reactor neutrino oscillation experiments with a Large Liquid Scintillator Detector. We consider two different scenarios for a measurement of the small mixing angle $\\theta_{13}$ with a mobile $\\bar{\

  11. Results on θ13 Neutrino Oscillations from Reactor Experiments

    Directory of Open Access Journals (Sweden)

    Kim Soo-Bong

    2014-03-01

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

  12. 77 FR 16098 - In the Matter of All Operating Boiling Water Reactor Licensees With Mark I and Mark II...

    Science.gov (United States)

    2012-03-19

    ... and containment failure. The leakage of hydrogen gas into the reactor buildings resulted in explosions... Regulatory Commission (NRC or Commission) authorizing operation of nuclear power plants in accordance with... Dai-ichi nuclear power plant site. The earthquake and tsunami produced widespread devastation...

  13. Experimental study of CHF enhancement using Fe{sub 3}O{sub 4} nanofluids at subcooled boiling region

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young Jae; Kam, Dong Hoon; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-05-15

    Two phase flow heat transfer mechanism is evaluated importantly in thermal hydraulic field due to big latent heat of liquid-vapor phase change. The mechanism of boiling and condensing plays an important role on removing heat for nuclear safety. Especially Critical Heat Flux(CHF) is very important for boiling heat transfer. Various methods to improve CHF characteristics are introduced, especially nanofluids are used for enhancing the CHF. Nanofluids is a colloidal suspension that nanoparticles are mixed with basic fluid. Normally the use of nanofluids as working fluid improves the flow boiling CHF characteristics. Lee et al. already researched the CHF characteristics using nanofluids. As exit quality increased from 0.07 to 0.74, CHF enhancement gradually decreased and approached zero. CHF enhancement was observed when exit quality was low and a DNB-like thermal crisis occurred. But CHF enhancement didn't occur for high exit quality, but LFD-type thermal crisis occurred. Because LFD phenomena are nearly unaffected by the surface conditions, CHF enhancement is not expected for annular flow with high exit quality. Kim et al. performed flow boiling CHF enhancement at subcooled region using alumina-water, zinc-oxide-water and diamond-water nanofluids. The CHF was enhanced by increasing wettability from nanoparticle deposition. CHF enhancement occurred in high mass flux (2000-2500 kg/m{sup 2}s), but CHF enhancement didn't occur in low mass flux (1500 kg/m{sup 2}s). Our experimental procedures are modified differently because the amount of nanoparticle deposition of each tube can be different by the several condition such as deposition time, mass flux and heat flux. So, the purpose of our experiment is to confirm how the CHF enhancement by nanofluids changes according to exit quality when same amount of nanoparticle is deposited at all tube.

  14. An overview of the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Cao, Jun

    2016-01-01

    The Daya Bay Reactor Neutrino Experiment discovered an unexpectedly large neutrino oscillation related to the mixing angle $\\theta_{13}$ in 2012. This finding paved the way to the next generation of neutrino oscillation experiments. In this article, we review the history, featured design, and scientific results of Daya Bay. Prospects of the experiment are also described.

  15. Planned reactor and beam experiments on Neutrino Oscillations

    Energy Technology Data Exchange (ETDEWEB)

    Goodman, Maury [Argonne National Lab, Argonne IL 60439 (United States)

    2009-08-15

    Current and future neutrino oscillation experiments are discussed with an emphasis on those that will measure or further limit the neutrino oscillation parameter {theta}{sub 13}. Some {nu}{sub e} disappearance experiments are being planned at nuclear reactors, and more ambitious {nu}{sub {mu}}{yields}{nu}{sub e} appearance experiments are being planned using accelerator beams.

  16. Enhanced heat transfer in confined pool boiling

    NARCIS (Netherlands)

    Rops, C.M.; Lindken, R.; Velthuis, J.F.M.; Westerweel, J.

    2009-01-01

    We report the results of an experimental investigation of the heat transfer during nucleate boiling on a spatially confined boiling surface. The heat flux as a function of the boiling surface temperature was measured in pool boiling pots with diameters ranging from 15 mm down to 4.5 mm. It was found

  17. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  18. First result from the Double Chooz reactor-neutrino experiment

    CERN Document Server

    Matsubara, Tsunayuki

    2012-01-01

    We report first results of a search for the non-zero neutrino mixing angle \\theta_{13} from the Double Chooz experiment. Double Chooz aims to measure the mixing angle based on anti-electron-neutrino disappearance as a consequence of neutrino oscillation. A new generation of anti-electron-neutrino detector having 10 m^3 fiducial volume is located 1 km from the two 4.25 GW_{th} reactors at the Chooz Power Plant in France. Physics data taking has been continuing since April 2011. A ratio of observed-to-predicted event rate of 0.944 +/- 0.016 (stat) +/- 0.040 (syst) was obtained in 101 days of detector running. Analyzing both the rate and their energy spectral shape, we found sin^{2}2\\theta_{13} = 0.086 +/- 0.041 (stat) +/- 0.030 (syst) at \\Delta m^2_{atm} = 2.4 x 10^{-3} eV^2.

  19. Double Chooz and a history of reactor θ13 experiments

    Science.gov (United States)

    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.

  20. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  1. New strategies of reloads design and models of control bars in boiling water reactors; Nuevas estrategias de diseno de recargas y de patrones de barras de control en reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Castillo M, J. A.; Ortiz S, J. J.; Perusquia del Cueto, R., E-mail: alejandro.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    In this work the results obtained when analyzing new strategies in the reload designs of nuclear fuel and models of control bars, for boiling water reactors are presented. The idea is to analyze the behaviour of the reactor during an operation cycle, when the heuristic rules are not used (commonly used by expert engineers in both designs). Specifically was analyzed the rule of low leak and the load strategy Control Cell Core for the design of a fuel reload. In a same way was analyzed the rule of prohibiting the use of the intermediate positions in the control bars, as well as the construction of bar models based on load strategies type Control Cell Core. In the first analysis a balance and transition cycle were used. For the second analysis only a transition cycle was used, firstly with the reloads designed in the first analysis and later on with reloads built by other methods. For the simulation of the different configurations proposed in both cases, was used the code Simulate-3. To obtain the designs in both studies, the heuristic techniques or neural networks and taboo search were used. The obtained results show that it can be omitted of some rules used in the ambit for the mentioned designs and even so to obtain good results. To carry out this investigation was used Dell work station under Li nux platform. (Author)

  2. Reactor Neutrino Flux Uncertainty Suppression on Multiple Detector Experiments

    CERN Document Server

    Cucoanes, Andi; Cabrera, Anatael; Fallot, Muriel; Onillon, Anthony; Obolensky, Michel; Yermia, Frederic

    2015-01-01

    This publication provides a coherent treatment for the reactor neutrino flux uncertainties suppression, specially focussed on the latest $\\theta_{13}$ measurement. The treatment starts with single detector in single reactor site, most relevant for all reactor experiments beyond $\\theta_{13}$. We demonstrate there is no trivial error cancellation, thus the flux systematic error can remain dominant even after the adoption of multi-detector configurations. However, three mechanisms for flux error suppression have been identified and calculated in the context of Double Chooz, Daya Bay and RENO sites. Our analysis computes the error {\\it suppression fraction} using simplified scenarios to maximise relative comparison among experiments. We have validated the only mechanism exploited so far by experiments to improve the precision of the published $\\theta_{13}$. The other two newly identified mechanisms could lead to total error flux cancellation under specific conditions and are expected to have major implications o...

  3. Numerical study of subcooled boiling phenomena using a component analysis code, CUPID

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ba-Ro; Lee, Yeon-Gun [Jeju National University, Jeju (Korea, Republic of)

    2015-10-15

    In this study, a couple of subcooled boiling experiments at high- (> 10 bar) and low-pressure (near atmospheric pressure) conditions are analyzed using a three-dimensional thermal-hydraulic component code, CUPID. And then the analysis results compared with the results using MARS-KS code. Subcooled boiling experiments at high- and low pressure conditions are analyzed using a three dimensional thermal-hydraulic component code, CUPID. The predictions of the CUPID code shows good agreement with Christenses's data and Bartolomey's data obtained at high pressure conditions. Subcooled boiling is encountered in many industrial applications in the power and process industry. In nuclear reactors, under certain conditions, subcooled boiling may be encountered in the core. The movement of bubbles generated by subcooled boiling affect the heat transfer characteristics and the pressure drop of the system. Thus some experimental and analysis using safety codes works have been already performed by previous investigators. It has been reported that the existing safety analysis codes have some weaknesses in predicting subcooled boiling phenomena at low pressure conditions. Thus, it is required to improve the predictive capability of thermal-hydraulic analysis codes on subcooled boiling phenomenon at low-pressure conditions. At low pressure condition, the CUPID code generally is overestimated prediction of the void fraction. Thus, we did selected submodels in the heat partitioning model by sensitivity analysis. Selected submodels of M{sub c}ase 4 are Kocamustafaogullari and Ishii correlation model of active nucleate site density, N' and Fritz correlation model of bubble departure diameter, d{sub Bd} . And then, case 5 - 8 are reanalysis using submodels of M{sub c}ase 4. The calculated void fraction is compared the default CUPID code model to the modified CUPID code model. As a result, average void fraction error was reduced from 0.081 to 0.011 and 0.128 to 0.024, 0

  4. Nucleate boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Saiz Jabardo, J.M. [Universidade da Coruna (Spain). Escola Politecnica Superior], e-mail: mjabardo@cdf.udc.es

    2009-07-01

    Nucleate boiling heat transfer has been intensely studied during the last 70 years. However boiling remains a science to be understood and equated. In other words, using the definition given by Boulding, it is an 'insecure science'. It would be pretentious of the part of the author to explore all the nuances that the title of the paper suggests in a single conference paper. Instead the paper will focus on one interesting aspect such as the effect of the surface microstructure on nucleate boiling heat transfer. A summary of a chronological literature survey is done followed by an analysis of the results of an experimental investigation of boiling on tubes of different materials and surface roughness. The effect of the surface roughness is performed through data from the boiling of refrigerants R-134a and R-123, medium and low pressure refrigerants, respectively. In order to investigate the extent to which the surface roughness affects boiling heat transfer, very rough surfaces (4.6 {mu}m and 10.5 {mu}m ) have been tested. Though most of the data confirm previous literature trends, the very rough surfaces present a peculiar behaviour with respect to that of the smoother surfaces (Ra<3.0 {mu}m). (author)

  5. Experiments on Void Fraction of CO2 Flow Boiling in a Horizontal Micro-fin Tube

    Science.gov (United States)

    Kondou, Chieko; Higashiiue, Shinya; Kuwahara, Ken; Koyama, Shigeru

    This paper deals with an experimental investigation on the void fraction of CO2 flow boiling in a horizontal micro-fin tube. The mean void fraction in the insulated 400 mm length sampling section, which is located next to the test evaporator, has measured by the quick closing valve method. The experimental data have been obtained in mass flux range of 200 to 455 kg/(m2s) and the refrigerant pressure range of 3.5 to 5.0 MPa. It is confirmed that the relation between void fraction and quality is affected by both mass flux and pressure. The experimental results are also compared with two previous correlations for horizontal smooth tubes, which are proposed by Butterworth and Smith. The present data satisfactorily agreed with Butterworth's correlation in the range of quality from 0.03 to 0.99. However, Smith's correlation is found to predict slightly higher than present data. As a trial, the empirical correlation of void fraction, based on the experimental slip ratios, is proposed.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Investigating the spectral anomaly with different reactor antineutrino experiments

    Science.gov (United States)

    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.

  8. Investigating the spectral anomaly with different reactor antineutrino experiments

    Directory of Open Access Journals (Sweden)

    C. Buck

    2017-02-01

    Full Text Available 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 U235 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 U235 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.

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

    Science.gov (United States)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Analysis of the rotation accident of assemblies in boiling water reactors; Analisis del accidente de rotacion de ensambles en reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Becerril-Gonzalez M, J. J. [Universidad Autonoma de Yucatan, Av. Industrias no contaminantes por Anillo Periferico Norte s/n, Apdo. Postal 150 Cordemex, Merida, Yucatan (Mexico); Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia de Cueto, R., E-mail: juanjosebecerril_1@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    For this work was analyzed the impact that would cause the load of a rotated fuel assembly in the behaviour of the core in the Cycle 14 of the Unit 1 of the nuclear power plant of Laguna Verde. To carry out this analysis the code Simulate-3 was used, with which was possible to analyze the behavior of the effective multiplication factor and the thermal limits (MAPRAT, MFLPD and MFLCPR). The rotation of fuel assemblies to 90, 180 and 270 grades was analyzed with regard to the design position, with 0, 1, 2 and 3 burnt cycles for these assemblies. The results show that the thermal limits remain inside the allowed values, therefore if this accident type happened the reactor could continue operating in a sure way. (Author)

  12. Indication for the disappearance of reactor electron antineutrinos in the Double Chooz experiment

    CERN Document Server

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

    2011-01-01

    The Double Chooz Experiment presents an indication of reactor electron antineutrino disappearance consistent with neutrino oscillations. A ratio of 0.944 $\\pm$ 0.016 (stat) $\\pm$ 0.040 (syst) observed to predicted events was obtained in 101 days of running at the Chooz Nuclear Power Plant in France, with two 4.25 GW$_{th}$ reactors. The results were obtained from a single 10 m$^3$ fiducial volume detector located 1050 m from the two reactor cores. The reactor antineutrino flux prediction used the Bugey4 measurement as an anchor point. The deficit can be interpreted as an indication of a non-zero value of the still unmeasured neutrino mixing parameter \\sang. Analyzing both the rate of the prompt positrons and their energy spectrum we find \\sang = 0.086 $\\pm$ 0.041 (stat) $\\pm$ 0.030 (syst), or, at 90% CL, 0.015 $<$ \\sang $\\ <$ 0.16.

  13. Inspection of Pool Boiling with Superhydrophilic and Superhydrophobic Coating

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    In conventional nuclear power plants, increasing critical heat flux (CHF) margin by converting existing parts is economically meaningful since it means overall energy production increase without building additional power plants. There were researches to enhance margin from the very beginning of the commercialization of nuclear power plants and many efforts have led to current model of plants, optimized for both safety and production efficiency. Examples are mixing vane which is actually applied to plants nowadays, using nanofluids to enhance heat transfer coefficient (HTC), trying porous surfaces and so on. Takata et al. studied effects of surface wettability by using hydrophobic coating and observed enhanced nucleate boiling at coated surface regions. Betz et al. experimented superhydrophilic (SHPi), superhydrophobic (SHPo), and superbiphilic surfaces. Results indicate heat transfer coefficient enhancement due to increase of nucleation sites by hydrophobic regions and constrained diameter of growing bubbles by hydrophilic regions. Although it would be rough to apply their concept to real reactor coolant surface wall, understanding the possibility of enhanced boiling is meaningful. In this paper, SHPi and SHPo coatings were applied to wire at traditional pool boiling experiment by Nukiyama. By observing altered CHF margin and nucleate boiling, the effects of each coating and their tendencies are discussed. SHPi, SHPo and bare wire's pool boiling was investigated and their boiling graphs were discussed. SHPi shows enhancement in CHF while SHPo's case is more complicated since there were variables like partial CHF or micro scale bubbles. Additional experiment could be comparing HTC, checking whether hydrophobic wire's nucleate boiling enhancement can exceed the decreased CHF margin. More sophisticated method to remove unwanted bubbles should be considered such as using degassed water.

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

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Enrico Sartori; Lori Scott

    2006-09-01

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

  15. Determination of local boiling in light water reactors by correlation of the neutron noise; Determination de l'ebullition locale dans les reacteurs a eau legere par correlation du bruit neutronique

    Energy Technology Data Exchange (ETDEWEB)

    Zwingelstein, G. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-07-01

    The power limit of swimming-pool type reactors depends on the phenomenon of the appearance of burn-out. In order to determine this limit we have attempted to detect the local boiling which usually occurs before the burn out. Local boiling has been simulated by an electrically heated plate placed in the core of the reactor Siloette. The study of local boiling, which is based on the properties of the correlation functions for the neutron noise of detectors placed in the core, shows that a privileged frequency occurs in the power spectrum of the noise. It is intended in the future to determine the influence of various parameters on this characteristic frequency. (author) [French] La limitation de la puissance des reacteurs nucleaires de type piscine est due au phenomene d'apparition de 'burn out'. Pour determiner cette limitation, nous nous sommes proposes dans ce rapport de detecter l'ebullition locale qui apparait generalement avant le 'burn out'. L'ebullition locale a ete simulee par une plaque chauffee electriquement et placee dans le coeur du reacteur SILOETTE. L'etude de l'ebullition locale, qui est basee sur les proprietes des fonctions de correlation du bruit neutronique de detecteurs places clans le coeur, fait apparaitre une frequence privilegiee dans le spectre de puissance du bruit. On envisage dans l'avenir, de determiner l'influence des divers parametres sur cette frequence caracteristique. (auteur)

  16. Results and Prospects from the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Higuera, A

    2016-01-01

    The Daya Bay reactor experiment has reported the most precise measurement of sin$^{2}2\\theta_{13}$ and $\\Delta m^{2}_{ee}$ by using a data set with the fully constructed design of 8 antineutrino detectors (ADs). We also report on a new independent measurement of sin$^{2}2\\theta_{13}$ from neutron capture on hydrogen, which confirms the results using gadolinium caputres. Several other analyses are also performed, including a measurements on the absolute reactor antineutrino flux and a search for light sterile neutrinos. Prospects for new analyses such as searching for CPT/LI violation at Daya Bay are ongoing.

  17. Optimization of operation schemes in boiling water reactors using neural networks; Optimizacion de esquemas de operacion en reactores de agua en ebullicion usando redes neuronales

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Pelta, D. A., E-mail: juanjose.ortiz@inin.gob.mx [Universidad de Granada, Escuela Superior de Ingenierias, Informatica y Telecomunicacion, C/Daniel Saucedo Aranda s/n, 18071 Granada (Spain)

    2012-10-15

    In previous works were presented the results of a recurrent neural network to find the best combination of several groups of fuel cells, fuel load and control bars patterns. These solution groups to each problem of Fuel Management were previously optimized by diverse optimization techniques. The neural network chooses the partial solutions so the combination of them, correspond to a good configuration of the reactor according to a function objective. The values of the involved variables in this objective function are obtained through the simulation of the combination of partial solutions by means of Simulate-3. In the present work, a multilayer neural network that learned how to predict some results of Simulate-3 was used so was possible to substitute it in the objective function for the neural network and to accelerate the response time of the whole system of this way. The preliminary results shown in this work are encouraging to continue carrying out efforts in this sense and to improve the response quality of the system. (Author)

  18. 应用于反应堆热工水力程序的核态沸腾传热关系式评价%Assessment of Nucleate Boiling Correlations Applied to Thermal-hydraulic Code of Reactor

    Institute of Scientific and Technical Information of China (English)

    李美琳; 林萌; 杨燕华; 张昊; 龚湛

    2015-01-01

    This article is set in development of reactor thermal-hydraulic analysis code-COSINE,studies variation characteristic of results of nucleate boiling correlations usually used for thermal-hydraulic analysis codes with variable changing,compares difference degree of different correlations and sensitivity in different ranges based on fuel rod wall heat transfer and external reactor vessel coolant conditions,in order to provide advice on setting of code user options and further experiment researches. It draws a conclusion that the condition of high superheat degree is mostly needed experimental demonstration and Chen,Schrock-Grossman1,Wright and Schrock-Grossman2 correlations are more suitable for calculating these conditions in reactor thermal-hydraulic analysis code.%本文以反应堆热工水力分析程序 COSINE 开发为背景,针对燃料棒和冷却剂换热及压力容器外部冷却时的核态沸腾两种特殊的工况,研究常用于计算热工水力程序的核态沸腾传热关系式的计算结果随影响参数的变化关系,比较不同范围内各关系式计算结果的差异程度和敏感性,为程序中用户选项的设置和进一步实验验证提供参考意见,研究表明高过热度工况最需进行实验验证,反应堆热工水力分析程序计算这两种工况下的核态沸腾传热更适宜选用 Chen、Schrock-Grossman1、Wright 和 Schrock-Grossman2公式。

  19. Training experience at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, J.W.; McCormick, R.P.; McCreery, H.I.

    1978-01-01

    The EBR-II Training Group develops, maintains,and oversees training programs and activities associated with the EBR-II Project. The group originally spent all its time on EBR-II plant-operations training, but has gradually spread its work into other areas. These other areas of training now include mechanical maintenance, fuel manufacturing facility, instrumentation and control, fissile fuel handling, and emergency activities. This report describes each of the programs and gives a statistical breakdown of the time spent by the Training Group for each program. The major training programs for the EBR-II Project are presented by multimedia methods at a pace controlled by the student. The Training Group has much experience in the use of audio-visual techniques and equipment, including video-tapes, 35 mm slides, Super 8 and 16 mm film, models, and filmstrips. The effectiveness of these techniques is evaluated in this report.

  20. Oscillate Boiling

    CERN Document Server

    Li, Fenfang; Nguyen, Dang Minh; Ohl, Claus-Dieter

    2016-01-01

    We report about an intriguing boiling regime occurring for small heaters embedded on the boundary in subcooled water. The microheater is realized by focusing a continuous wave laser beam to about $10\\,\\mu$m in diameter onto a 165\\,nm-thick layer of gold, which is submerged in water. After an initial vaporous explosion a single bubble oscillates continuously and repeatably at several $100\\,$kHz. The microbubble's oscillations are accompanied with bubble pinch-off leading to a stream of gaseous bubbles into the subcooled water. The self-driven bubble oscillation is explained with a thermally kicked oscillator caused by the non-spherical collapses and by surface pinning. Additionally, Marangoni stresses induce a recirculating streaming flow which transports cold liquid towards the microheater reducing diffusion of heat along the substrate and therefore stabilizing the phenomenon to many million cycles. We speculate that this oscillate boiling regime may allow to overcome the heat transfer thresholds observed dur...

  1. Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors; Etude experimentale des instabilites hydrodynamiques survenant dans les reacteurs nucleaires a ebullition

    Energy Technology Data Exchange (ETDEWEB)

    Fabreca, S. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-10-01

    The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [French] II a ete effectue une etude experimentale, en boucle hors-pile, des oscillations hydrodynamiques survenant dans les reacteurs a ebullition. L'etude a ete effectuee a la pression atmospherique et a une pression voisine de 8 atmospheres dans des canaux chauffes electriquement a puissance imposee constante et uniforme. Dans les essais a 8 atmospheres le canal etait un tube circulaire de diametre interieur 6 mm environ. A 1 atmosphere le canal etait de section annulaire 10 * 20 mm avec un tube interieur chauffant et un tube exterieur en pyrex. Le fonctionnement etait possible

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

    Science.gov (United States)

    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

  3. Submersion-Subcritical Safe Space (S{sup 4}) reactor

    Energy Technology Data Exchange (ETDEWEB)

    King, Jeffrey C. [Institute for Space and Nuclear Power Studies, Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131 (United States)]. E-mail: isnps@unm.edu; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131 (United States)]. E-mail: mgenk@unm.edu

    2006-09-15

    The Submersion-Subcritical Safe Space (S{sup 4}) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S{sup 4} reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S{sup 4} reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S{sup 4} reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S{sup 4} reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively.

  4. Results from the Daya Bay Reactor Neutrino Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Tsang, K.V. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States); An, F.P. [Institute of High Energy Physics, Beijing (China); An, Q. [University of Science and Technology of China, Hefei (China); Bai, J.Z. [Institute of High Energy Physics, Beijing (China); Balantekin, A.B.; Band, H.R. [University of Wisconsin, Madison, WI (United States); Beriguete, W.; Bishai, M. [Brookhaven National Laboratory, Upton, NY (United States); Blyth, S. [National United University, Miao-Li (China); Brown, R.L. [Brookhaven National Laboratory, Upton, NY (United States); Cao, G.F.; Cao, J. [Institute of High Energy Physics, Beijing (China); Carr, R. [California Institute of Technology, Pasadena, CA (United States); Chan, W.T. [Brookhaven National Laboratory, Upton, NY (United States); Chang, J.F. [Institute of High Energy Physics, Beijing (China); Chang, Y. [National United University, Miao-Li (China); Chasman, C. [Brookhaven National Laboratory, Upton, NY (United States); Chen, H.S. [Institute of High Energy Physics, Beijing (China); Chen, H.Y. [Institute of Physics, National Chiao-Tung University, Hsinchu (China); Chen, S.J. [Nanjing University, Nanjing (China); and others

    2014-01-15

    The Daya Bay Reactor Neutrino Experiment was designed to achieve a sensitivity on the value of sin{sup 2}2θ{sub 13} 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 sin{sup 2}2θ{sub 13} 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 sin{sup 2}2θ{sub 13}=0.089±0.010(stat.)±0.005(syst.), which is the most precise measurement of sin{sup 2}2θ{sub 13} to date.

  5. The Radon Monitoring System in Daya Bay Reactor Neutrino Experiment

    CERN Document Server

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

    2016-01-01

    We developed a highly sensitive, reliable and portable automatic system (H$^{3}$) to monitor the radon concentration of the underground experimental halls of the Daya Bay Reactor Neutrino Experiment. H$^{3}$ is able to measure radon concentration with a statistical error less than 10\\% in a 1-hour measurement of dehumidified air (R.H. 5\\% at 25$^{\\circ}$C) with radon concentration as low as 50 Bq/m$^{3}$. This is achieved by using a large radon progeny collection chamber, semiconductor $\\alpha$-particle detector with high energy resolution, improved electronics and software. The integrated radon monitoring system is highly customizable to operate in different run modes at scheduled times and can be controlled remotely to sample radon in ambient air or in water from the water pools where the antineutrino detectors are being housed. The radon monitoring system has been running in the three experimental halls of the Daya Bay Reactor Neutrino Experiment since November 2013.

  6. Paramagnetism and improved upconversion luminescence properties of NaYF4:Yb,Er/NaGdF4 nanocomposites synthesized by a boiling water seed-mediated route

    Science.gov (United States)

    Yang, Chao-Qing; Li, Ao-Ju; Guo, Wei; Tian, Peng-Hua; Yu, Xiao-Long; Liu, Zhong-Xin; Cao, Yang; Sun, Zhong-Liang

    2016-03-01

    In a route boiling water served as reaction medium, a stoichiometric amount of rare-earth compound and fluoride are put into this system to form α-NaYF4:Yb, Er nuclei. Then prepared sample is heated at elevated temperature to improve the fluorescence intensity, and next a NaGdF4 shell grows on the surface of NaYF4 nuclei. NaYF4:Yb,Er/NaGdF4 core-shell structured upconversion nanoparticles (CSUCNPs) have been successfully synthesized by above route. The use of boiling water decreases the cubic-to-hexagonal phase transition temperature of NaYF4:Yb,Er to 350°C and increases its upconversion (UC) luminescence intensity. A heterogeneous NaGdF4 epitaxially growing on the surface of Ln3+-doped NaYF4 not only improves UC luminescence, but also creates a paramagnetic shell, which can be used as contrast agents in magnetic resonance imaging (MRI). The solution of CSUCNPs shows bright green UC fluorescence under the excitation at 980 nm in a power density only about 50 mW·cm-2. A broad spectrum with a dominant resonance at g of about 2 is observed by the electron paramagnetic resonance (EPR) spectrum of CSUCNPs. Above properties suggest that the obtained CSUCNPs could be potential candidates for dual-mode optical/magnetic bioapplications.

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

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J.H. [Basic Science Research Institute, Dongshin University, Naju 58245 (Korea, Republic of); Jang, H.I. [Department of Fire Safety, Seoyeong University, Gwangju 61268 (Korea, Republic of); Choi, W.Q. [Department of Physics & Astronomy, Seoul National University, Seoul 08826 (Korea, Republic of); Choi, Y. [Department of Physics, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Jang, J.S. [GIST College, Gwangju Institute of Science and Technology, Gwangju 61005 (Korea, Republic of); Jeon, E.J. [Institute for Basic Science, Daejeon 34047 (Korea, Republic of); Department of Physics and Astronomy, Sejong University, Seoul 05006 (Korea, Republic of); Joo, K.K.; Kim, B.R. [Institute for Universe & Elementary Particles, Chonnam National University, Gwangju 61186 (Korea, Republic of); Kim, H.S. [Department of Physics and Astronomy, Sejong University, Seoul 05006 (Korea, Republic of); Kim, J.Y. [Institute for Universe & Elementary Particles, Chonnam National University, Gwangju 61186 (Korea, Republic of); Kim, S.B.; Kim, S.Y. [Department of Physics & Astronomy, Seoul National University, Seoul 08826 (Korea, Republic of); Kim, W. [Department of Physics, Kyungpook National University, Daegu 41566 (Korea, Republic of); Kim, Y.D. [Institute for Basic Science, Daejeon 34047 (Korea, Republic of); Ko, Y.J. [Department of Physics, Chung-Ang University, Seoul 06974 (Korea, Republic of); Lee, J.K. [Department of Physics & Astronomy, Seoul National University, Seoul 08826 (Korea, Republic of); Lim, I.T. [Institute for Universe & Elementary Particles, Chonnam National University, Gwangju 61186 (Korea, Republic of); Pac, M.Y., E-mail: pac@dsu.kr [Basic Science Research Institute, Dongshin University, Naju 58245 (Korea, Republic of); Park, I.G. [Department of Physics, Gyeongsang National University, Jinju 52828 (Korea, Republic of); Park, J.S. [Department of Physics & Astronomy, Seoul National University, Seoul 08826 (Korea, Republic of); and others

    2016-02-21

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

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    Cuta, Judith M.; Krotiuk, William

    1988-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. A review of film boiling at cryogenic temperatures.

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Recent Results from Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Hu, Bei-Zhen

    2015-01-01

    The Daya Bay reactor neutrino experiment announced the discovery of a non-zero value of \\sin^22\\theta_{13} with significance better than 5 \\sigma in 2012. The experiment is continuing to improve the precision of \\sin^22\\theta_{13} and explore other physics topics. In this talk, I will show the current oscillation and mass-squared difference results which are based on the combined analysis of the measured rates and energy spectra of antineutrino events, an independent measurement of \\theta_{13} using IBD events where delayed neutrons are captured on hydrogens, and a search for light sterile neutrinos.

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

    Science.gov (United States)

    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.

  16. Cyclic voltammetry: a tool to quantify 2,4,6-trichloroanisole in aqueous samples from cork planks boiling industrial process.

    Science.gov (United States)

    Peres, António M; Freitas, Patrícia; Dias, Luís G; Sousa, Mara E B C; Castro, Luís M; Veloso, Ana C A

    2013-12-15

    Chloroanisoles, namely 2,4,6-trichloroanisole, are pointed out as the primary responsible of the development of musty off-flavours in bottled wine, due to their migration from cork stoppers, which results in huge economical losses for wine industry. A prevention step is the detection of these compounds in cork planks before stoppers are produced. Mass spectrometry gas chromatography is the reference method used although it is far beyond economical possibilities of the majority of cork stoppers producers. In this work, a portable cyclic voltammetry approach was used to detect 2,4,6-trichloroanisole extracted from natural cork planks to the aqueous phase during the cork boiling industrial treatment process. Analyses were carried out under ambient conditions, in less than 15 min with a low use of solvent and without any sample pre-treatment. The proposed technique had detection (0.31±0.01 ng/L) and quantification (0.95±0.05 ng/L) limits lower than the human threshold detection level. For blank solutions, without 2,4,6-trichloroanisole addition, a concentration in the order of the quantification limit was estimated (1.0±0.2 ng/L), which confirms the satisfactory performance of the proposed methodology. For aqueous samples from the industrial cork planks boiling procedure, intra-day repeatabilities were lower than 3%, respectively. Also, 2,4,6-trichloroanisole contents in the aqueous samples determined by this novel approach were in good agreement with those obtained by GC-MS (correlation coefficient equal to 0.98), confirming the satisfactory accuracy of the proposed methodology. So, since this novel approach is a fast, low-cost, portable and user-friendly method, it can be an alternative and helpful tool for in-situ industrial applications, allowing accurate detection of releasable 2,4,6-trichloroanisole in an earlier phase of cork stoppers production, which may allow implementing more effective cork treatments to reduce or avoid future 2,4,6-trichloroanisole

  17. Reactor electron antineutrino disappearance in the Double Chooz experiment

    CERN Document Server

    Abe, Y; Anjos, J C dos; 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; Goger-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; Rohling, M; Roncin, R; Roth, S; Sakamoto, Y; Santorelli, R; Sato, F; Schonert, 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; Stuken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Svoboda, R; Terao, K; Tonazzo, A; Toups, M; Thi, H H Trinh; Valdiviesso, G; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yermia, F; Zimmer, V

    2012-01-01

    The Double Chooz experiment has observed 8,249 candidate electron antineutrino events in 227.93 live days with 33.71 GW-ton-years (reactor power x detector mass x livetime) exposure using a 10.3 cubic meter fiducial volume detector located at 1050 m from the reactor cores of the Chooz nuclear power plant in France. The expectation in case of theta13 = 0 is 8,937 events. The deficit is interpreted as evidence of electron antineutrino disappearance. From a rate plus spectral shape analysis we find sin^2 2{\\theta}13 = 0.109 \\pm 0.030(stat) \\pm 0.025(syst). The data exclude the no-oscillation hypothesis at 99.9% CL (3.1{\\sigma}).

  18. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    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. Study of in-reactor creep of vanadium alloy in the HFIR RB-12J experiment

    Energy Technology Data Exchange (ETDEWEB)

    Strain, R.V.; Konicek, C.F.; Tsai, H. [Argonne National Lab., IL (United States)

    1996-10-01

    Biaxial creep specimens will be included in the HFIR RB-12J experiment to study in-reactor creep of the V-4Cr-4Ti alloy at {approx}500{degrees}C and 5 dpa. These specimens were fabricated with the 500-kg, heat (832665) material and pressurized to attain 0, 50, 100, 150, and 200 MPa mid-wall hoop stresses during the irradiation.

  1. Analysis of the DHCE experiment in the position A10 of the ATR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, I.C.; Smith, D.L.; Tsai, H. [Argonne National Lab., IL (United States)

    1997-08-01

    Calculations were performed to assess the possibility of performing DHCE experiments in mixed spectrum fission reactors. Calculated values of key parameters were compared with limit values for each quantity. The values calculated were: He-4 production from the {sup 6}Li(n,t){sup 4}He reaction, tritium leakage, required tritium concentration in lithium, initial tritium charge per capsule, and helium to dpa ratio after 10 dpa of irradiation.

  2. Reactor scram experience for shutdown system reliability analysis. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Edison, G.E.; Pugliese, S.L.; Sacramo, R.F.

    1976-06-01

    Scram experience in a number of operating light water reactors has been reviewed. The date and reactor power of each scram was compiled from monthly operating reports and personal communications with the operating plant personnel. The average scram frequency from ''significant'' power (defined as P/sub trip//P/sub max greater than/ approximately 20 percent) was determined as a function of operating life. This relationship was then used to estimate the total number of reactor trips from above approximately 20 percent of full power expected to occur during the life of a nuclear power plant. The shape of the scram frequency vs. operating life curve resembles a typical reliability bathtub curve (failure rate vs. time), but without a rising ''wearout'' phase due to the lack of operating data near the end of plant design life. For this case the failures are represented by ''bugs'' in the plant system design, construction, and operation which lead to scram. The number of scrams would appear to level out at an average of around three per year; the standard deviations from the mean value indicate an uncertainty of about 50 percent. The total number of scrams from significant power that could be expected in a plant designed for a 40-year life would be about 130 if no wearout phase develops near the end of life.

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

    Science.gov (United States)

    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.

  4. Film boiling on vertical surfaces.

    Science.gov (United States)

    Suryanarayana, N. V.; Merte, H., Jr.

    1972-01-01

    Film boiling of a saturated liquid on a vertical surface is analyzed to determine the local heat-transfer rates as a function of height and heater-surface superheat. Experiments show that the laminar-flow model is inadequate. A turbulent-vapor-flow model is used, and the influence of the interfacial oscillations is incorporated on a semiempirical basis. Measurements of local film boiling were obtained with a transient technique using saturated liquid nitrogen.

  5. Estimation of Control Rod Worth in a VVER-1000 Reactor using DRAGON4 and DONJON4

    Directory of Open Access Journals (Sweden)

    Saadatian-derakhshandeh Farahnaz

    2014-07-01

    Full Text Available One of the main issues in safety and control systems design of power and research reactors is to prevent accidents or reduce the imposed hazard. Control rod worth plays an important role in safety and control of reactors. In this paper, we developed a justifiable approach called D4D4 to estimate the control rod worth of a VVER-1000 reactor that enables to perform the best estimate analysis and reduce the conservatism that utilize DRAGON4 and DONJON4. The results are compared with WIMS-D4/CITATION to show the effectiveness and superiority of the developed package in predicting reactivity worth of the rod and also other reactor physics parameters of the VVER-1000 reactor. The results of this study are in good agreement with the plant's FSAR.

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

    Energy Technology Data Exchange (ETDEWEB)

    An, F.P. [Institute of Modern Physics, East China University of Science and Technology, Shanghai (China); Institute of High Energy Physics, Beijing (China); Balantekin, A.B. [University of Wisconsin, Madison, WI (United States); Band, H.R. [Department of Physics, Yale University, New Haven, CT (United States); University of Wisconsin, Madison, WI (United States); Beriguete, W.; Bishai, M. [Brookhaven National Laboratory, Upton, NY (United States); Blyth, S. [Department of Physics, National Taiwan University, Taipei (China); National United University, Miao-Li, Taiwan (China); Brown, R.E. [Brookhaven National Laboratory, Upton, NY (United States); Butorov, I. [Joint Institute for Nuclear Research, Dubna, Moscow Region (Russian Federation); Cao, G.F.; Cao, J. [Institute of High Energy Physics, Beijing (China); Carr, R. [California Institute of Technology, Pasadena, CA (United States); Chan, Y.L. [Chinese University of Hong Kong (Hong Kong); Chang, J.F. [Institute of High Energy Physics, Beijing (China); Chang, L. [Institute of Physics, National Chiao-Tung University, Hsinchu, Taiwan (China); Chang, Y. [National United University, Miao-Li, Taiwan (China); Chasman, C. [Brookhaven National Laboratory, Upton, NY (United States); Chen, H.S. [Institute of High Energy Physics, Beijing (China); Chen, H.Y. [Institute of Physics, National Chiao-Tung University, Hsinchu, Taiwan (China); Chen, Q.Y. [Shandong University, Jinan (China); Chen, S.J. [Nanjing University, Nanjing (China); and others

    2015-02-11

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Recent Results from the Daya Bay Reactor Neutrino Experiment

    Science.gov (United States)

    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 absolute measurement of the rate and energy spectrum of reactor antineutrinos will also be presented.

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

    Energy Technology Data Exchange (ETDEWEB)

    Briere, E.; Larrauri, D.; Olive, J. [Electricite de France, Chatou (France)

    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.

  12. Hydraulic Experiment for Simulative Assemblies of Blanket Assembly and Np Transmutation Assembly of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    CHENG; Dao-xi; QI; Xiao-guang; ZHAI; Wei-ming; YANG; Bing; ZHOU; Ping

    2013-01-01

    The out-of reactor hydraulic experiment of fast reactor assembly is one of the important experiments in the process of the development of the fast reactor assembly.In this experiment,the size of the throttling element in the foot of the assembly is decided which is fit for the flow division in the reactor and the

  13. Heat transfer measurements on smooth and finned tubes in a standard apparatus for boiling experiments. Waermeuebergangsmessungen an Glatt- und Rippenrohren in einer Standardapparatur fuer Siedeversuche

    Energy Technology Data Exchange (ETDEWEB)

    Fath, W.

    1986-07-28

    The use of Rankine cycles with refrigerants or other organic working fluids for waste heat recovery has been a matter of discussion lately. The evaporating pressure is higher in these systems than in conventional application of finned tubes, so the problems of heat transfer in nucleate boiling on finned tubes are investigated here. Measurements were made on smooth and finned tubes and on surface-treated smooth and finned tubes. Refrigerant R-22 was used for the experiments; the measurements cover a wide range of pressures. (HAG).

  14. Experimental research on heat transfer to liquid sodium and its incipient boiling wall superheat in an annulus

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Liquid sodium is mainly used as a cooling fluid in the liquid metal fast breeder reactor (LMFBR), whose heat transfer, whether convective heat transfer or boiling heat transfer, is different from that of water. So it is important for both normal and accidental operations of LMFBR to perform experimental research on heat transfer to liquid sodium and its boiling heat transfer. This study deals with heat transfer with high temperature (300-700℃) and low Pe number (20~70) and heat transfer with low temperature (250~270℃) and high Pe number (125~860), and its incipient boiling wall superheat in an annulus. Research on heat transfer involves theoretical research and experiments on heat transfer to liquid sodium. It also focuses on the theoretical analysis and experimental research on its incipient boiling wall superheat at positive pressure in an annulus. Semiempirical correlations were obtained and they were well coincident with the experimental data.

  15. Heat transfer 1982; Proceedings of the Seventh International Conference, Technische Universitaet Muenchen, Munich, West Germany, September 6-10, 1982. Volume 4 - General papers: Pool boiling, flow boiling, measuring techniques

    Science.gov (United States)

    Grigull, U.; Straub, J.; Hahne, E.; Stephan, K.

    Papers are presented on nucleate pool boiling in a microgravity environment; the influence of diameter on nucleate boiling outside tubes; burnout during flow across a small cylinder influenced by parallel cylinders; and the application of the swollen polymer technique to the study of heat transfer on film-cooled surfaces. Other topics include homogeneous nucleation in transient boiling; the heat transfer characteristics of the two-phase closed thermosyphon (wickless heat pipe); boiling heat transfer in annular flow; and the effect of rod bundle geometry (even-odd) on burnout. Attention is also given to wall temperature fluctuation of the evaporating tube at the dryout region, to the flow boiling of ethanol/cyclohexane mixtures, and to capillary flows around hemispherical bubbles. For individual items see A83-42758 to A83-42781

  16. Duality of boiling systems and uncertainty phenomena

    Institute of Scientific and Technical Information of China (English)

    柴立合; 彭晓峰; 王补宣

    2000-01-01

    Interactions among dry patches at high heat flux are theoretically analyzed. The high heat flux boiling experiments on metal plate wall with different materials and thickness are correspondingly conducted. The duality of boiling system, i.e. hydrodynamic performance and self-organized performance is identified. A unified explanation of hydrodynamic models and dry patches models is given. The scatter and uncertainty in boiling data can be mainly attributed to the intrinsic duality, but not the sole surface effects. The present experimental results explain why the deviation point at high flux boiling is seen only on occasion and why the self-organization of dry patches is often ignored in available literature.

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

    Science.gov (United States)

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

    2015-04-01

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

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

    Science.gov (United States)

    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.

  19. Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

    Directory of Open Access Journals (Sweden)

    Chang Dong Shin

    2014-01-01

    Full Text Available For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ13, providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reactor neutrino experiments, more precise measurements of θ12,  Δm122, and mass hierarchy will be explored. The precise measurement of θ13 would be crucial for measuring the CP violation parameters at accelerators. Therefore, reactor neutrino physics will assist in the complete understanding of the fundamental nature and implications of neutrino masses and mixing. In this paper, we investigated several characteristics of RENO-50, which is a future medium-baseline reactor neutrino oscillation experiment, by using the GloBES simulation package.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

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

  1. Sodium boiling detection in LMFBRs by acoustic-neutronic cross correlation

    Energy Technology Data Exchange (ETDEWEB)

    Wright, S.A.

    1977-01-01

    The acoustic and neutronic noise signals caused by boiling are the signals primarily considered likely to detect sodium boiling in an LMFBR. Unfortunately, these signals may have serious signal-to-noise problems due to strong background noise sources. Neutronic-acoustic cross correlation techniques are expected to provide a means of improving the signal-to-noise ratio. This technique can improve the signal-to-noise ratio because the neutronic and acoustic signals due to boiling are highly correlated near the bubble repetition frequency, while the background noise sources are expected to be uncorrelated (or at most weakly correlated). An experiment was designed to show that the neutronic and acoustic noise signals are indeed highly correlated. The experiment consisted of simulating the void and pressure effects of local sodium boiling in the core of a zero-power reactor (ARK). The analysis showed that the neutronic and acoustic noise signals caused by boiling are almost perfectly correlated in a wide frequency band about the bubble repetition frequency. The results of the experiments were generalized to full-scale reactors to compare the inherent effectiveness of the methods which use the neutronic or acoustic signals alone with a hybrid method, which cross correlates the neutronic and acoustic signals. It was concluded that over a zone of the reactor where the void coefficient is sufficiently large (approximately 85 percent the core volume), the cross correlation method can provide a more rapid detection system for a given signal-to-noise ratio. However, where the void coefficient is small, one must probably rely on the acoustic method alone.

  2. Study on model of onset of nucleate boiling in natural circulation with subcooled boiling using unascertained mathematics

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Tao [Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China)]. E-mail: zhoutao@mail.tsinghua.edu.cn; Wang Zenghui [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China); Yang Ruichang [Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China)

    2005-10-01

    Experiment data got from onset of nucleate boiling (ONB) in natural circulation is analyzed using unascertained mathematics. Unitary mathematics model of the relation between the temperature and onset of nucleate boiling is built up to analysis ONB. Multiple unascertained mathematics models are also built up with the onset of natural circulation boiling equation based on the experiment. Unascertained mathematics makes that affirmative results are a range of numbers that reflect the fluctuation of experiment data more truly. The fluctuating value with the distribution function F(x) is the feature of unascertained mathematics model and can express fluctuating experimental data. Real status can be actually described through using unascertained mathematics. Thus, for calculation of ONB point, the description of unascertained mathematics model is more precise than common mathematics model. Based on the unascertained mathematics, a new ONB model is developed, which is important for advanced reactor safety analysis. It is conceivable that the unascertained mathematics could be applied to many other two-phase measurements as well.

  3. Pool Boiling of Hydrocarbon Mixtures on Water

    Energy Technology Data Exchange (ETDEWEB)

    Boee, R.

    1996-09-01

    In maritime transport of liquefied natural gas (LNG) there is a risk of spilling cryogenic liquid onto water. The present doctoral thesis discusses transient boiling experiments in which liquid hydrocarbons were poured onto water and left to boil off. Composition changes during boiling are believed to be connected with the initiation of rapid phase transition in LNG spilled on water. 64 experimental runs were carried out, 14 using pure liquid methane, 36 using methane-ethane, and 14 using methane-propane binary mixtures of different composition. The water surface was open to the atmosphere and covered an area of 200 cm{sup 2} at 25 - 40{sup o}C. The heat flux was obtained by monitoring the change of mass vs time. The void fraction in the boiling layer was measured with a gamma densitometer, and a method for adapting this measurement concept to the case of a boiling cryogenic liquid mixture is suggested. Significant differences in the boil-off characteristics between pure methane and binary mixtures revealed by previous studies are confirmed. Pure methane is in film boiling, whereas the mixtures appear to enter the transitional boiling regime with only small amounts of the second component added. The results indicate that the common assumption that LNG will be in film boiling on water because of the high temperature difference, may be questioned. Comparison with previous work shows that at this small scale the results are influenced by the experimental apparatus and procedures. 66 refs., 76 figs., 28 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  6. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  7. Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor

    CERN Document Server

    Sinev, V V

    2009-01-01

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

  8. Lessons Learned about Liquid Metal Reactors from FFTF Experience

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance, (France); Snoj, Luka [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana, (Slovenia); Tarchalski, Mikolaj [National Centre for Nuclear Research, ulica Andrzeja Soltana 7, 05-400 Otwock (Swierk), (Poland); Dewynter-Marty, Veronique [CEA, DEN, DANS, DRSN, SIREN, LESCI, Saclay, F-91191 Gif sur Yvette, (France); Malouch, Fadhel [CEA, DEN, DANS, DM2S, SERMA, Saclay, F-91191 Gif sur Yvette, (France)

    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)

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

    CERN Document Server

    Bilenky, S M

    2016-01-01

    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.

  11. 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)

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  13. Reactor core design and characteristics of the Fugen

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Mitsuo; Kowata, Yasuki; Sugawara, Satoru; Deshimaru, Takehide

    1988-03-01

    The heavy water moderated, boiling light water cooled pressure tube type reactor Fugen uses plutonium-uranium mixed oxide as a fuel. Heavy water as the moderator and the light water of coolant are separated by the pressure tubes and calandria tubes. Thereby, the reactor core is heterogenes compared with that of LWRs. This paper describes the development of reactor core design procedure based on the feature of the Fugen type reactor, the feasibility test and the validity of nuclear and thermalhydraulic design based on the operating experience.

  14. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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…

  16. Application of a diffusion-reaction kinetic model for the removal of 4-chlorophenol in continuous tank reactors.

    Science.gov (United States)

    Murcia, M D; Gómez, M; Bastida, J; Hidalgo, A M; Montiel, M C; Ortega, S

    2014-08-01

    A continuous tank reactor was used to remove 4-chlorophenol from aqueous solutions, using immobilized soybean peroxidase and hydrogen peroxide. The influence of operational variables (enzyme and substrate concentrations and spatial time) on the removal efficiency was studied. By using the kinetic law and the intrinsic kinetic parameters obtained in a previous work with a discontinuous tank reactor, the mass-balance differential equations of the transient state reactor model were solved and the theoretical conversion values were calculated. Several experimental series were used to obtain the values of the remaining model parameters by numerical calculation and using an error minimization algorithm. The model was checked by comparing the results obtained in some experiments (not used for the determination of the parameters) and the theoretical ones. The good concordance between the experimental and calculated conversion values confirmed that the design model can be used to predict the transient behaviour of the reactor.

  17. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    Science.gov (United States)

    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

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

    Science.gov (United States)

    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.

  19. Biodegradation of Jet Fuel-4 (JP-4) in Sequencing Batch Reactors

    Science.gov (United States)

    1992-06-01

    Specific Objectives of This Proposal Are: 1. To assess the ability of C. resinae , P. chrysosporium and selected bacterial consortia to degrade individual...chemical components of JP-4. 2. To develop a sequencing batch reactor that utilizes C. resinae to degrade chemical components of JP-4 in contaminated...Angeles Air Force Base. The study of the effectiveness of the two fungal systems (P. chrysosporium and and C. resinae ) is still in progress. Initial

  20. Evaporation, Boiling and Bubbles

    Science.gov (United States)

    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…

  1. Mass hierarchy sensitivity of medium baseline reactor neutrino experiments with multiple detectors

    CERN Document Server

    Wang, Hongxin; Li, Yu-Feng; Cao, Guofu; Chen, Shenjian

    2016-01-01

    We report the neutrino mass hierarchy (MH) sensitivity of medium baseline reactor neutrino experiments with multiple detectors. Sensitivity of determining the MH can be significantly improved by adding a near detector and combining both the near and far detectors. The size of the sensitivity improvement is related to accuracy of the individual mass-splitting measurements and requires strict control on the relative energy scale uncertainty of the near and far detectors. We study the impact of both baseline and target mass of the near detector on the combined sensitivity. A figure-of-merit is defined to optimize the baseline and target mass of the near detector and the optimal selections are $\\sim$13~km and $\\sim$4~kton respectively for a far detector with the 20~kton target mass and 52.5~km baseline. As typical examples of future medium baseline reactor neutrino experiments, the optimal location and target mass of the near detector are selected for JUNO and RENO-50. Finally, we discuss distinct effects of the ...

  2. The entropy balance for boiling flow

    Energy Technology Data Exchange (ETDEWEB)

    Collado, Francisco-Javier E-mail: fjk@posta.unizar.es

    2001-10-01

    Subcooled forced convection boiling of water is recognized as one of the best means of accommodating the very high heat fluxes that plasma facing components of fusion reactors have to withstand. The boiling curve, giving the wall temperature in function of the applied flux and flow conditions, is essential for the design of such cooling configurations. In this paper, a new entropy balance for subcooled boiling flow, which allows the wall temperature to be obtained, is presented and successfully compared with experimental data from the Joint US-EURATOM R and D Program. The derivation of this entropy balance is based on a new strict application of the Reynolds theorem to multiphase flows recently proposed by the author.

  3. Uncertainties analysis of fission fraction for reactor antineutrino experiments using DRAGON

    CERN Document Server

    Ma, X B; Chen, Y X; Zhong, W L; An, F P

    2014-01-01

    Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulation to predict reactor rates. First, DRAGON was developed to calculate the fission rates of the four most important isotopes in fissions,235U,238U,239Pu and141Pu, and it was validated for PWRs using the Takahama benchmark. The fission fraction calculation function was validated through comparing our calculation results with MIT's results. we calculate the fission fraction of the Daya Bay reactor core, and compare its with those calculated by the commercial reactor simulation program SCIENCE, which is used by the Daya Bay nuclear power plant, and the results was consist with each other. The uncertainty of the antineutrino flux by the fission fraction was studied, and the uncertainty of the antineutrino flux by the fission fraction simulation is 0.6% per core for Daya Bay antineutrino experiment.

  4. Hydraulic characterization of an activated sludge reactor with recycling system by tracer experiment and analytical models.

    Science.gov (United States)

    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.

  5. Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Guentay, S. [ed.] [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-12-01

    The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs.

  6. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  7. Heat Transfer in Boiling Dilute Emulsion with Strong Buoyancy

    Science.gov (United States)

    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

  8. Air purification in a reverse-flow reactor: Model simulations vs. experiments

    OpenAIRE

    Beld, van de, L.; Westerterp, K.R.

    1996-01-01

    The behavior of a reverse-flow reactor was studied for the purification of polluted air by catalytic combustion. A heterogeneous one-dimensional model was extended with a heat balance for the reactor wall. An overall heat transport term is included to account for the small heat losses in radial direction. The calculations are compared to experimental data without using fit parameters. The agreement between simulations and experiments is generally good. Discrepancies can be explained mainly by...

  9. Aspects of subcooled boiling

    Energy Technology Data Exchange (ETDEWEB)

    Bankoff, S.G. [Northwestern Univ., Evanston, IL (United States)

    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.

  10. Rate-Only analysis with reactor-off data in the Double Chooz experiment

    CERN Document Server

    Novella, P

    2013-01-01

    Among ongoing reactor-based experiments, Double Chooz is unique in obtaining data when the reactor cores are brought down for maintenance. These reactor-off data allow for a clean measurement of the backgrounds of the experiment, thus being of uppermost importance for the theta13 oscillation analysis. While the oscillation results published by the collaboration in 2011 and 2012 rely on background models derived from reactor-on data, in this talk we present an independent study based on the handle provided by 7.53 days of reactor-off data. A global fit to both theta13 and the total background is performed by analyzing the observed neutrino rate as a function of the non-oscillated expected rate for different reactor power conditions. The result presented in this talk is fully consistent with the one already published by Double Chooz. As they both yield almost the same precision, this work stands as a prove of the reliability of the background estimates and the oscillation analysis of the experiment.

  11. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    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.

  12. Precision Neutrino Oscillation Physics with an Intermediate Baseline Reactor Neutrino Experiment

    CERN Document Server

    Choubey, S; Piai, M; Choubey, Sandhya

    2003-01-01

    We discuss the physics potential of intermediate $L \\sim 20 \\div 30$ km baseline experiments at reactor facilities, assuming that the solar neutrino oscillation parameters $\\Delta m^2_{\\odot}$ and $\\theta_{\\odot}$ lie in the high-LMA solution region. We show that such an intermediate baseline reactor experiment can determine both $\\Delta m^2_{\\odot}$ and $\\theta_{\\odot}$ with a remarkably high precision. We perform also a detailed study of the sensitivity of the indicated experiment to $\\Delta m^2_{\\rm atm}$, which drives the dominant atmospheric $\

  13. Integrated photocatalytic-biological reactor for accelerated 2,4,6-trichlorophenol degradation and mineralization.

    Science.gov (United States)

    Zhang, Yongming; Sun, Xia; Chen, Lujun; Rittmann, Bruce E

    2012-02-01

    An integrated photocatalytic-biological reactor (IPBR) was used for accelerated degradation and mineralization of 2,4,6-trichlorophenol (TCP) through simultaneous, intimate coupling of photocatalysis and biodegradation in one reactor. Intimate coupling was realized by circulating the IPBR's liquid contents between a TiO(2) film on mat glass illuminated by UV light and honeycomb ceramics as biofilm carriers. Three protocols-photocatalysis alone (P), biodegradation alone (B), and integrated photocatalysis and biodegradation (photobiodegradation, P&B)-were used for degradation of different initial TCP concentrations. Intimately coupled P&B also was compared with sequential P and B. TCP removal by intimately coupled P&B was faster than that by P and B alone or sequentially coupled P and B. Because photocatalysis relieved TCP inhibition to biodegradation by decreasing its concentration, TCP biodegradation could become more important over the full batch P&B experiments. When phenol, an easy biodegradable compounds, was added to TCP in order to promote TCP mineralization by means of secondary utilization, P&B was superior to P and B in terms of mineralization of TCP, giving 95% removal of chemical oxygen demand. Cl(-) was only partially released during P experiments (24%), and this corresponded to its poor mineralization in P experiments (32%). Thus, intimately coupled P&B in the IPBR made it possible obtain the best features of each: rapid photocatalytic transformation in parallel with mineralization of photocatalytic products.

  14. Partial Nucleate Pool Boiling at Low Heat Flux: Preliminary Ground Test for SOBER-SJ10

    Science.gov (United States)

    Wu, Ke; Li, Zhen-Dong; Zhao, Jian-Fu; Li, Hui-Xiong; Li, Kai

    2016-05-01

    Focusing on partial nucleate pool boiling at low heat flux, SOBER-SJ10, one of 27 experiments of the program SJ-10, has been proposed to study local convection and heat transfer around an isolated growing vapor bubble during nucleate pool boiling on a well characterized flat surface in microgravity. An integrated micro heater has been developed. By using a local pulse overheating method in the experimental mode of single bubble boiling, a bubble nucleus can be excited with accurate spatial and temporal positioning on the top-side of a quartz glass substrate with a thickness of 2 mm and an effective heating area of 4.5 mm in diameter, and then grows under an approximate constant heat input provided by the main heater on the back-side of the substrate. Ten thin film micro-RTDs are used for local temperature measurements on the heating surface underneath the growing bubble. Normal pool boiling experiments can also be carried out with step-by-step increase of heating voltage. A series of ground test of the flight module of SOBER-SJ10 have been conducted. Good agreement of the measured data of single phase natural convection with the common-used empirical correlation warrants reasonable confidence in the data. It is found that the values of the incipience superheat of pool boiling at different subcooling are consistent with each others, verifying that the influence of subcooling on boiling incipience can be neglected. Pool boiling curves are also obtained, which shows great influence of subcooling on heat transfer of partial nucleate pool boiling, particularly in lower heat flux.

  15. Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M. [Oak Ridge National Lab., TN (United States); Suto, T. [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)]|[Oak Ridge National Lab., TN (United States)

    1996-10-01

    ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k{sub eff} of 1. 0040{+-}0.0005.

  16. A Matrix Method of Analyzing the Thermodynamic System of Advance Boiling Water Reactor Nuclear Power Unit%先进型沸水堆核电机组热经济性矩阵分析方法

    Institute of Scientific and Technical Information of China (English)

    冉鹏; 李庚生; 廖丹; 朱伟平

    2010-01-01

    根据先进型沸水堆(advance boiling water reactor,ABWR)核电机组热力系统的结构特点,基于热力系统等效热降分析方法和矩阵方法,确定其主、辅系统的划分原则以及辅助汽水成分划分原则,对先进型沸水堆各种汽水成分进行归并处理,构建表达规则的先进型沸水堆核电机组汽水分布方程填写规则,推导出适合先进型沸水堆核电机组热力系统热经济性分析的通用矩阵方法,并给出该类型核电机组辅助汽水成分对热经济性影响的表达方式.该矩阵全面反映了先进犁沸水堆核电机组热力系统主系统和各种辅助系统对机组热经济性的影响状况,每个子矩阵物理意义明确、规律性强,可使先进型沸水堆核电机组热力系统的整体计算和局部分析变得清晰、简单,适合于计算机程序化,并通过实例对该方法进行了验证.

  17. Solar Neutrino Oscillation Parameters in Experiments with Reactor Anti-Neutrinos

    CERN Document Server

    Choubey, Sandhya

    2004-01-01

    We review the current status of the solar neutrino oscillation parameters. We discuss the conditions under which measurements from future solar neutrino experiments would determine the oscillation parameters precisely. Finally we expound the potential of long baseline reactor anti-neutrino experiments in measuring the solar neutrino oscillation parameters.

  18. Boiling in porous media; Ebullition en milieux poreux

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-11

    This conference day of the French society of thermal engineers was devoted to the analysis of heat transfers and fluid flows during boiling phenomena in porous media. This book of proceedings comprises 8 communications entitled: `boiling in porous medium: effect of natural convection in the liquid zone`; `numerical modeling of boiling in porous media using a `dual-fluid` approach: asymmetrical characteristic of the phenomenon`; `boiling during fluid flow in an induction heated porous column`; `cooling of corium fragment beds during a severe accident. State of the art and the SILFIDE experimental project`; `state of knowledge about the cooling of a particulates bed during a reactor accident`; `mass transfer analysis inside a concrete slab during fire resistance tests`; `heat transfers and boiling in porous media. Experimental analysis and modeling`; `concrete in accidental situation - influence of boundary conditions (thermal, hydric) - case studies`. (J.S.)

  19. Fluid phase thermodynamics : I) nucleate pool boiling of oxygen under magnetically enhanced gravity and II) superconducting cavity resonators for high-stability frequency references and precision density measurements of helium-4 gas

    Science.gov (United States)

    Corcovilos, Theodore Allen

    Although fluids are typically the first systems studied in undergraduate thermodynamics classes, we still have only a rudimentary phenomenological understanding of these systems outside of the classical and equilibrium regimes. Two experiments will be presented. First, we present progress on precise measurements of helium-4 gas at low temperatures (1 K-5 K). We study helium because at low densities it is an approximately ideal gas but at high densities the thermodynamic properties can be predicted by numerical solutions of Schroedinger's equation. By utilizing the high resolution and stability in frequency of a superconducting microwave cavity resonator we can measure the dielectric constant of helium-4 to parts in 109, corresponding to an equivalent resolution in density. These data will be used to calculate the virial coefficients of the helium gas so that we may compare with numerical predictions from the literature. Additionally, our data may allow us to measure Boltzmann's constant to parts in 108, a factor of 100 improvement over previous measurements. This work contains a description of the nearly-completed apparatus and the methods of operation and data analysis for this experiment. Data will be taken by future researchers.The second experiment discussed is a study of nucleate pool boiling. To date, no adequate quantitative model exists of this everyday phenomenon. In our experiment, we vary one parameter inaccessible to most researchers, gravity, by applying a magnetic force to our test fluid, oxygen. Using this technique, we may apply effective gravities of 0-80 times Earth's gravitational acceleration (g). In this work we present heat transfer data for the boiling of oxygen at one atmosphere ambient pressure for effective gravity values between 1g and 16g . Our data describe two relationships between applied heat flux and temperature differential: at low heat flux the system obeys a power law and at high heat flux the behavior is linear. We find that the

  20. Integral Circulation Experiment: Thermal-hydraulic simulator of a heavy liquid metal reactor

    Science.gov (United States)

    Tarantino, M.; Agostini, P.; Benamati, G.; Coccoluto, G.; Gaggini, P.; Labanti, V.; Venturi, G.; Class, A.; Liftin, K.; Forgione, N.; Moreau, V.

    2011-08-01

    In the frame of the IP-EUROTRANS (6th Framework Program EU), domain DEMETRA, ENEA was involved in the Work Package 4.5 " Large Scale Integral Test", devoted to characterize a relevant portion of a sub-critical ADS reactor block (core, internals, heat exchanger, cladding for fuel elements) in steady state, transient and accidental conditions. More in details ENEA assumed the commitment to perform an integral experiment aiming to reproduce the primary flow path of the " European Transmutation Demonstrator (ETD)" pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE). This experimental activity, called " Integral Circulation Experiment (ICE)", has been implemented merging the efforts of several research institutes, among which, besides ENEA, FZK, CRS4 and University of Pisa, allowing to design an appropriate test section to be installed in the CIRCE facility. The goal of the experiments is therefore to demonstrate the technological feasibility of a heavy liquid metal (HLM) nuclear system pool-type in a relevant scale (1 MW), investigating the related thermal-hydraulic behaviour (heat source and heat exchanger coupling, primary system and downcomer coupling, gas trapping into the main stream, thermal stratification in the pool, forced and mixed convection in rod bundle) under both steady state and transient conditions. Moreover the preliminary as well as the planned experiments aims to address performance and reliability tests of some prototypical components, such as heat source, heat exchanger, chemistry control system. The paper reports a detailed description of the experiment, the design performed for the test section and its main components as well as the preliminary experimental results carried out in the first experimental campaign run on the CIRCE pool, which consists of a full power steady state test. The preliminary experimental results carried out have demonstrate the proper design of the test section trough the experiment goals as well as the HLM

  1. Integral Circulation Experiment: Thermal-hydraulic simulator of a heavy liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tarantino, M., E-mail: mariano.tarantino@enea.it [ENEA UTIS, C.R. Brasimone, 40032 Camugnano, BO (Italy); Agostini, P.; Benamati, G.; Coccoluto, G.; Gaggini, P.; Labanti, V.; Venturi, G. [ENEA UTIS, C.R. Brasimone, 40032 Camugnano, BO (Italy); Class, A.; Liftin, K. [KIT, Forschungszentrum Karlsruhe, IKET, P.O. Box 3640, D-76021 Karlsruhe (Germany); Forgione, N. [Universita di Pisa, DIMNP, Via Diotisalvi 2, 56126 Pisa (Italy); Moreau, V. [CRS4, Loc. Piscina Manna, Edificio 1, 09010 Pula (Italy)

    2011-08-31

    In the frame of the IP-EUROTRANS (6th Framework Program EU), domain DEMETRA, ENEA was involved in the Work Package 4.5 'Large Scale Integral Test', devoted to characterize a relevant portion of a sub-critical ADS reactor block (core, internals, heat exchanger, cladding for fuel elements) in steady state, transient and accidental conditions. More in details ENEA assumed the commitment to perform an integral experiment aiming to reproduce the primary flow path of the 'European Transmutation Demonstrator (ETD)' pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE). This experimental activity, called 'Integral Circulation Experiment (ICE)', has been implemented merging the efforts of several research institutes, among which, besides ENEA, FZK, CRS4 and University of Pisa, allowing to design an appropriate test section to be installed in the CIRCE facility. The goal of the experiments is therefore to demonstrate the technological feasibility of a heavy liquid metal (HLM) nuclear system pool-type in a relevant scale (1 MW), investigating the related thermal-hydraulic behaviour (heat source and heat exchanger coupling, primary system and downcomer coupling, gas trapping into the main stream, thermal stratification in the pool, forced and mixed convection in rod bundle) under both steady state and transient conditions. Moreover the preliminary as well as the planned experiments aims to address performance and reliability tests of some prototypical components, such as heat source, heat exchanger, chemistry control system. The paper reports a detailed description of the experiment, the design performed for the test section and its main components as well as the preliminary experimental results carried out in the first experimental campaign run on the CIRCE pool, which consists of a full power steady state test. The preliminary experimental results carried out have demonstrate the proper design of the test section trough the

  2. Reactor

    Science.gov (United States)

    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.

  3. Tests of Lorentz and CPT Violation in the Medium Baseline Reactor Antineutrino Experiment

    CERN Document Server

    Li, Yu-Feng

    2014-01-01

    Tests of Lorentz and CPT violation in the medium baseline reactor antineutrino experiment are presented in the framework of the Standard Model Extension (SME). Both the spectral distortion and sidereal variation are employed to derive the limits of Lorentz violation (LV) coefficients. We do the numerical analysis of the sensitivity of LV coefficients by taking the Jiangmen Underground Neutrino Observatory (JUNO) as an illustration, which can improve the sensitivity by more than two orders of magnitude compared with the current limits from reactor antineutrino experiments.

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

    Science.gov (United States)

    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.

  5. A New Jacobian Matrix Method for Assessing Similarity between Critical Experiments and Real Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    For a metal fueled Sodium-cooled Fast Reactor (SFR), innovative reactors such as Prototype Gen-IV Sodium cooled Fast Reactor (PGSFR), unfortunately, experiment data from an operating reactor are unavailable because there are few operating reactors in the world. Hence, a critical experiment is the only way to obtain meaningful experiment data for the target core. However, there is a considerable geometrical difference between the critical assembly for a critical experiment and the target core. The neutron characteristics of a system are influenced by the geometrical difference. A number of researches have been performed to confirm the similarity between a critical experiment and a real reactor using a conventional representativity factor. The conventional representativity factor defined as S{sup T}{sub E}US{sub R} √S{sup T}{sub E}US{sub E}, √S{sup T}{sub R}US{sub R} provides insight of similarity between two sensitivity vectors for cross sections, but it did not provide a quantitative value. Hence, up to now, the influence of geometrical difference to the reactivity is believed to be negligible. In this paper, a new Jacobian matrix method is proposed to provide a quantitative error for geometrical differences between two systems. In this method, reactivity of the critical assembly is decomposed into phenomenon-based reactivity and geometry-based reactivity. The reactivity is then transformed into a target core geometry using a Jacobian matrix. Then, non-linearity of two different systems can be derived by comparing the transformed reactivity with the original reactivity of the target core. The maximum error of the transformed reactivity can be used as an additional uncertainty of the geometrical difference.

  6. High flux film and transition boiling

    Science.gov (United States)

    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 the 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.

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

    Science.gov (United States)

    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. Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Higgins, J.; Nasta, K.

    1997-01-01

    The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs.

  9. 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.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. 78 FR 58575 - Review of Experiments for Research Reactors

    Science.gov (United States)

    2013-09-24

    ... NRC Library at http://www.nrc.gov/reading-rm/adams.html . To begin search, select ``ADAMS Public... innovations, ] congressional actions, or other events. Currently, guidance applicable to experiments...

  11. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  12. Technology Selection for Offshore Underwater Small Modular Reactors

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1 a lead–bismuth fast reactor based on the Russian SVBR-100; (2 a novel organic cooled reactor; (3 an innovative superheated water reactor; (4 a boiling water reactor based on Toshiba's LSBWR; and (5 an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80% with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

  13. Innovative reactor development strategy-focusing on generation 4

    Energy Technology Data Exchange (ETDEWEB)

    Matsui, Kazuaki [The Inst. of Applied Energy, Tokyo (Japan)

    2002-12-01

    Generation IV International Forum (GIF) has been established at the USDOE in 2000. Japan joined the GIF in July 2001 and now ten countries have joined to develop the nuclear power plant aiming at its realization in 2030's, which should be economically competitive with natural gas combined cycle power plant, proliferation-resistant, lower waste and safer system. These international activities claim for more efficient and accountable efforts of the next-generation reactor development. 94 reactor types of four kinds (water, gas, liquid metal and non classical) have been proposed from 12 countries and 20% proposals are from Japan. USNERAC (Nuclear Energy Research Advisory Committee) have selected three main indices to established the load map; sustainable development, safety and credibility and economics. The author reviewed the significance of GIF activities as a member of the team and commented how to relate them with Japanese internal activities. (T. Tanaka)

  14. Experience of IEA-R1 research reactor spent fuel transportation back to United States

    Energy Technology Data Exchange (ETDEWEB)

    Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Operacao do Reator IEAR-R1m]. E-mail: frajndli@net.ipen.br; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div.de Engenharia do Nucleo]. E-mail: perrotta@net.ipen.br; Maiorino, Jose Rubens [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Diretoria de Reatores]. E-mail: maiorino@net.ipen.br; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores]. E-mail: ajsoares@net.ipen.br

    1998-07-01

    IPEN/CNEN-SP is sending the IEA-R1 Research Reactor spent fuels from USA origin back to this country. This paper describes the experience in organizing the negotiations, documents and activities to perform the transport. Subjects as cask licensing, transport licensing and fuel failure criteria for transportation are presented. (author)

  15. Proceedings of a seminar on the potential for LMFBR boiling detection by acoustic/neutronic monitoring, Argonne, Illinois, April 8--9, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Carey, W.M.; Albrecht, R.W.

    1976-06-01

    A seminar involving ten technical presentations by principal investigators was held to assess the current scope of ERDA-sponsored programs to determine the feasibility of sodium-boiling detection in LMFBRs and to establish areas in need of additional research and development. The consensus was that (1) feasibility of boiling detection by acoustic, neutronic, and acoustic/neutronic monitors has been demonstrated in U.S. and European programs; (2) additional research and development is needed in areas of reactor noise, cavitation, and the effects of noncondensible gases on sound source levels and transmission; (3) the role of acoustic/neutronic monitors from the standpoint of reactor surveillance rather than reactor safety is a viable approach to be adapted; and, in particular (4) a need exists for an operational LMFBR demonstration system. Each paper has been separately abstracted and indexed. (DG)

  16. High Heat Flux Burnout in Subcooled Flow Boiling

    Institute of Scientific and Technical Information of China (English)

    G.P.Celata; M.Cumo; 等

    1995-01-01

    The paper reports the results of an experimental research carried out at the Heat transfer divison of the Energy Department,C.R.Casaccia,on the thermal hydraulic characterization of subcooled flow boiling CHF under typical conditions of thermonuclear fusion reactors.I.e.high liquid velocity and subcooling.The experiment was carried out exploring the following parameters:channel diameter(from 2.5to 8.0 mm),heated length(10 and 15cm) ,liquid velocity (from 2 to 40m/s),exit pressure(from atmospheric to 5.0 MPa),inlet temperature(from 30 to 80℃),channel orientation (vertical and horizontal),A maximum CHF value of 60.6MW/m2 has been obtained under the following conditions:Tin-30°,p=2.5MPa,u=40m/s,D=2.5mm(smooth channel) Turbulence promoters(helically coiled wires)have been employed to further enhance the CHF attainable with subcooled flow boiling.Helically coiled wires allow an increase of 50% of the maximum CHF obtained with smooth channels.

  17. The Lma MSW Solution of the Solar Neutrino Problem, Inverted Neutrino Mass Hierarchy and Reactor Neutrino Experiments

    CERN Document Server

    Petcov, S T

    2002-01-01

    In the context of three-neutrino oscillations, we study the possibility of using antineutrinos from nuclear reactors to explore the $10^-4 eV^2 < \\Delta m^2_{\\odot} \\ltap 8\\times 10^-4 eV^2$ region of the LMA MSW solution of the solar neutrino problem. The KamLAND experiment is not expected to determine $\\Delta m^2_{\\odot}$ if the latter happens to lie in the indicated region. By analysing both the total event rate suppression and the energy spectrum distortion caused by $\\bar{\

  18. Investigation of Enhanced Boiling Heat Transfer from Porous Surfaces

    Institute of Scientific and Technical Information of China (English)

    LinZhiping; MaTongze; 等

    1994-01-01

    Experimental investigations of boiling heat transfer from porous surfaces at atmospheric pressure were performne.The porous surfaces are plain tubes coverd with metal screens.V-shaped groove tubes covered with screens,plain tubes sintered with screens.and V-shaped groove tubes sintered with screens,The experimental results show that sintering metal screens around spiral V-shaped groove tubes can greatly improve the boiling heat transfer,The boiling hystesis was observed in the experiment.This paper discusses the mechanism of the boiling heat transfer from those kinds of porous surfaces stated above.

  19. A Study of the Influence of Solid Particles on Boiling Hysteresis

    Institute of Scientific and Technical Information of China (English)

    M.H.Shi; J.Ma

    1992-01-01

    Experiments have been performed to determine the effects on boiling hysteresis of locally fluidized particles contained in a liquid that serves as coolant for electronic equipment.The results show that Iocally fluidized particles can diminish boiling hysteresis.

  20. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. High Precision Measurements of $\\theta_{\\odot}$ in Solar and Reactor Neutrino Experiments

    CERN Document Server

    Bandyopadhyay, A; Goswami, S; Petcov, S T; Bandyopadhyay, Abhijit; Choubey, Sandhya; Goswami, Srubabati

    2004-01-01

    We discuss the possibilities of high precision measurement of the solar neutrino mixing angle $\\theta_\\odot \\equiv \\theta_{12}$ in solar and reactor neutrino experiments. The improvements in the determination of $\\sin^2\\theta_{12}$, which can be achieved with the expected increase of statistics and reduction of systematic errors in the currently operating solar and KamLAND experiments, are summarised. The potential of LowNu $\

  2. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past Incidents, Experiments, and Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

    2016-04-17

    In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

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

    Science.gov (United States)

    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.

  5. A new MC-based method to evaluate the fission fraction uncertainty at reactor neutrino experiment

    CERN Document Server

    Ma, X B; Chen, Y X

    2016-01-01

    Uncertainties of fission fraction is an important uncertainty source for the antineutrino flux prediction in a reactor antineutrino experiment. A new MC-based method of evaluating the covariance coefficients between isotopes was proposed. It was found that the covariance coefficients will varying with reactor burnup and which may change from positive to negative because of fissioning balance effect, for example, the covariance coefficient between $^{235}$U and $^{239}$Pu changes from 0.15 to -0.13. Using the equation between fission fraction and atomic density, the consistent of uncertainty of fission fraction and the covariance matrix were obtained. The antineutrino flux uncertainty is 0.55\\% which does not vary with reactor burnup, and the new value is about 8.3\\% smaller.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. The High Energy Neutrino Nuisance at a Medium Baseline Reactor Experiment

    CERN Document Server

    Ciuffoli, Emilio; Zhang, Xinmin

    2012-01-01

    10 years from now medium baseline reactor experiments will attempt to determine the neutrino mass hierarchy from the differences (RL+PV) between the extrema of the Fourier transformed neutrino spectra. Recently Qian et al. have claimed that this goal may be impeded by the strong dependence of the difference parameter RL+PV on the reactor neutrino flux and on slight variations of Delta M^2_32. We demonstrate that this effect results from a spurious dependence of the difference parameter on the very high energy (8+ MeV) tail of the reactor neutrino spectrum. This dependence is spurious because the high energy tail depends upon decays of exotic isotopes and is insensitive to the mass hierarchy. An energy-dependent weight in the Fourier transform not only eliminates this spurious dependence but in fact increases the chance of correctly determining the hierarchy.

  8. Manual Calibration System for Daya Bay Reactor Neutrino Experiment

    Institute of Scientific and Technical Information of China (English)

    HUANG; Han-xiong; RUAN; Xi-chao; REN; Jie; LV; Yin-long; FAN; Cheng-jun; CHEN; Yan-nan; WANG; Zhao-hui; ZHOU; Zu-ying; HOU; Long; ZHANG; Jia-wen; ZHANG; Yin-hong; YU; Chao-ju; HE; Wei; ZHOU; Bin

    2012-01-01

    <正>The neutrino mixing angle θ13 with a significance of 7.7 standard deviations has been published by the Daya Bay anti-neutrino experiment collaboration in 2012. To understand the non-uniformity and the energy non-linearity of the anti-neutrino detector (AD), a calibration campaign for the AD1 with the Manual Calibration System (MCS) has been finished. The aim of this calibration plan is to deploy the calibration sources to any positions inside the Inner Acrylic Vessel (IAV), to study detail properties of AD.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

  10. Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, Tomaz [Reactor Physics Department, Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)]. E-mail: tomaz.zagar@ijs.si; Bozic, Matjaz [Nuklearna elektrarna Krsko, Vrbina 12, 8270 Krsko (Slovenia); Ravnik, Matjaz [Reactor Physics Department, Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)

    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 ({gamma} emitting) radioactive nuclides in the concrete were found to be {sup 133}Ba, {sup 60}Co and {sup 152}Eu. 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 Jozef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. {sup 133}Ba, {sup 41}Ca) are not included in the IAEA and EU basic safety standards.

  11. Determining the neutrino mass hierarchy with INO, T2K, NOvA and reactor experiments

    Science.gov (United States)

    Ghosh, Anushree; Thakore, Tarak; Choubey, Sandhya

    2013-04-01

    The relatively large measured value of θ 13 has opened up the possibility of determining the neutrino mass hierarchy through earth matter effects. Amongst the current accelerator-based experiments only NOvA has a long enough baseline to observe earth matter effects. However, NOvA is plagued with uncertainty on the knowledge of the true value of δ CP, and this could drastically reduce its sensitivity to the neutrino mass hierarchy. The earth matter effect on atmospheric neutrinos on the other hand is almost independent of δ CP. The 50 kton magnetized Iron CALorimeter at the India-based Neutrino Observatory (ICAL@INO) will be observing atmospheric neutrinos. The charge identification capability of this detector gives it an edge over others for mass hierarchy determination through observation of earth matter effects. We study in detail the neutrino mass hierarchy sensitivity of the data from this experiment simulated using the NUANCE based generator developed for ICAL@INO and folded with the detector resolutions and efficiencies obtained by the INO collaboration from a full Geant4-based detector simulation. The data from ICAL@INO is then combined with simulated data from T2K, NOvA, Double Chooz, RENO and Daya Bay experiments and a combined sensitivity study to the mass hierarchy is performed. With 10 years of ICAL@INO data combined with T2K, NOvA and reactor data, one could get about 2.3 σ-5.7 σ discovery of the neutrino mass hierarchy, depending on the true value of sin2 θ 23 [0.4-0.6], sin2 2 θ 13 [0.08-0.12] and δ CP [0-2 π].

  12. Probing new physics scenarios in accelerator and reactor neutrino experiments

    Science.gov (United States)

    Di Iura, A.; Girardi, I.; Meloni, D.

    2015-06-01

    We perform a detailed combined fit to the {{\\bar{ν }}e}\\to {{\\bar{ν }}e} disappearence data of the Daya Bay experiment and the appearance {{ν }μ }\\to {{ν }e} and disappearance {{ν }μ }\\to {{ν }μ } data of the Tokai to Kamioka (T2K) one in the presence of two models of new physics affecting neutrino oscillations, namely a model where sterile neutrinos can propagate in a large compactified extra dimension and a model where non-standard interactions (NSI) affect the neutrino production and detection. We find that the Daya Bay ⨁ T2K data combination constrains the largest radius of the compactified extra dimensions to be R≲ 0.17 μm at 2σ C.L. (for the inverted ordering of the neutrino mass spectrum) and the relevant NSI parameters in the range O({{10}-3})-O({{10}-2}), for particular choices of the charge parity violating phases.

  13. Numerical Investigation of Boiling

    Science.gov (United States)

    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.

  14. Structural integrity and management of aging in internal components of BWR reactors; Integridad estructural y manejo del envejecimiento en componentes internos de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C.R. [Instituto Nacional de Investigaciones Nucleares, Km 36.5 Carretera Mexico, Toluca Salazar Edo. de Mexico (Mexico)]. E-mail: craj@nuclear.inin.mx

    2004-07-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

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

    KAUST Repository

    Inaoka, Hajime

    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. © 2013 Elsevier B.V. All rights reserved.

  16. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  17. Systematic simulation of a tubular recycle reactor on the basis of pilot plant experiments

    Energy Technology Data Exchange (ETDEWEB)

    Paar, H.; Narodoslawsky, M.; Moser, A. (Technische Univ., Graz (Austria). Inst. fuer Biotechnologie, Mikrobiologie und Abfalltechnologie)

    1990-10-10

    Systematic simulatiom may decisively help in development and optimization of bioprocesses. By applying simulation techniques, optimal use can be made of experimental data, decreasing development costs and increasing the accuracy in predicting the behavior of an industrial scale plant. The procedure of the dialogue between simulation and experimental efforts will be exemplified in a case study. Alcoholic fermentation of glucose by zymomonas mobilis bacteria in a gasified turbular recycle reactor was studied first by systematic simulation, using a computer model based solely on literature data. On the base of the results of this simulation, a 0.013 m{sup 3} pilot plant reactor was constructed. The pilot plant experiments, too, were based on the results of the systematic simulation. Simulated and experimental data were well in agreement. The pilot plant experiments reiterated the trends and limits of the process as shown by the simulation results. Data from the pilot plant runs were then used to improve the simulation model. This improved model was subsequently used to simulate the performances of an industrial scale plant. The results of this simulation are presented. They show that the alcohol fermentation in a tubular recycle reactor is potentially advantageous to other reactor configurations, especially to continuous stirred tanks. (orig.).

  18. Hydrogen/Oxygen Reactions at High Pressures and Intermediate Temperatures: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    A series of experimental and numerical investigations into hydrogen oxidation at high pressures and intermediate temperatures has been conducted. The experiments were carried out in a high pressure laminar flow reactor at 50 bar pressure and a temperature range of 600–900 K. The equivalence ratio......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time.......A series of experimental and numerical investigations into hydrogen oxidation at high pressures and intermediate temperatures has been conducted. The experiments were carried out in a high pressure laminar flow reactor at 50 bar pressure and a temperature range of 600–900 K. The equivalence ratio......, the mechanism is used to simulate published data on ignition delay time and laminar burning velocity of hydrogen. The flow reactor results show that at reducing, stoichiometric, and oxidizing conditions, conversion starts at temperatures of 750–775 K, 800–825 K, and 800–825 K, respectively. In oxygen atmosphere...

  19. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    Science.gov (United States)

    Farnum, Eugene H.; Shikama, Tatsuo; Narui, Minoru; Sagawa, Tsutomu; Scarborough, Kent

    1996-02-01

    An experiment to measure radiation-induced electrical degradation (RIED) in a sapphire sample and in three MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260°C to a fluence of 3 × 1024 n/m 2 ( E > 1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m. Even though the results for the sapphire sample are somewhat ambiguous because of an unexplained offset current of about 0.6 μA substantial degradation was not observed in the sapphire: instead, radiation-induced conductivity (RIC) seemed to decrease slightly during the experiment. Substantial increase in leakage current, that increased with applied electric field, occurred in the MgO-insulated cables. This increased conductivity disappeared when the reactor was shut down and sample temperature returned to ambient. However, the physical degradation apparently remained in the material while the reactor was off because restarting the irradiation brought the conductivity back to its previous, degraded, reactor-on value. This effect is different from the RIED effect reported by Hodgson but is similar to previous results reported by Shikama et al. Considerable data were taken to determine the sample temperature and leakage currents during the irradiation.

  20. HFO1234ze(E) And HFC134a Flow Boiling Inside a 4mm Horizontal Smooth Tube

    OpenAIRE

    Longo, Giovanni A.; Mancin, Simone; Righetti, Giulia; Zilio, Claudio

    2016-01-01

    Nowadays, the substitution of HFC134a with low GWP refrigerants is one of the most important challenge for refrigeration and air conditioning. The possible substitutes include natural refrigerants, such as HC600 (Butane) and HC600a (Isobutane), and also synthetic refrigerants, such as HFO1234yf and HFO1234ze(E). The HC refrigerants exhibit very low GWP, 3 and 4 HC600a and HC600 respectively, good thermodynamic and transport properties, and pressure and volumetric performance very similar to H...

  1. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  2. Effect of feeding time on the performance of a sequencing batch reactor treating a mixture of 4-CP and 2,4-DCP.

    Science.gov (United States)

    Sahinkaya, Erkan; Dilek, Filiz B

    2007-06-01

    This paper investigated the biodegradation kinetics of 4-chlorophenol (4-CP) and 2,4-dichlorophenol (2,4-DCP) separately in batch reactors and mixed in sequencing batch reactors (SBRs). Batch reactor experiments showed that both 4-CP and 2,4-DCP began to inhibit their own degradation at 53 and 25 mg l(-1), respectively, and that the Haldane equation gave a good fit to the experimental data because r(2) values were higher than 0.98. The maximum specific degradation rates (q(m)) were 130.3 and 112.4 mg g(-1) h for 4-CP and 2,4-DCP, respectively. The values of the half saturation (K(s)) and self-inhibition constants (K(i)) were 34.98 and 79.74 mg l(-1) for 4-CP, and 13.77 and 44.46 mg l(-1) for 2,4-DCP, respectively. The SBR was fed with a mixture of 220 mg l(-1) of 4-CP, 110 mg l(-1) of 2,4-DCP, and 300 mg l(-1) of peptone as biogenic substrate at varying feeding periods (0-8h) to evaluate the effect of feeding time on the performance of the SBR. During SBR operation, in addition to self-inhibition, 4-CP degradation was strongly and competitively inhibited by 2,4-DCP. The inhibitory effects were particularly pronounced during short feeding periods because of higher chlorophenol peak concentrations in the reactor. The competitive inhibition constant (K(ii)) of 2,4-DCP on 4-CP degradation was 0.17 mg l(-1) when the reactor was fed instantaneously (0 h feeding). During longer feedings, increased removal/loading rates led to lower chlorophenol peak concentrations at the end of feeding. Therefore, in multi-substrate systems feeding time plus reaction time should be determined based on both degradation kinetics and substrate interaction. During degradation, the meta cleavage of 4-chlorocatechol resulted in accumulation of a yellowish color because of the formation of 5-chloro-2-hydroxymuconic semialdehyde (CHMS), which was further metabolized. Isolation and enrichment of the chlorophenols-degrading culture suggested Pseudomonas sp. and Pseudomonas stutzeri to be the

  3. Characteristics of critical heat flux under rolling condition for flow boiling in vertical tube

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Jin-Seok, E-mail: hjscd@snu.ac.kr [Seoul National University, 599 Gwanak-Ro, Gwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Yeon-Gun, E-mail: yeongun2@snu.ac.kr [Seoul National University, 599 Gwanak-Ro, Gwanak-Gu, Seoul 151-742 (Korea, Republic of); Park, Goon-Cherl, E-mail: parkgc@snu.ac.kr [Seoul National University, 599 Gwanak-Ro, Gwanak-Gu, Seoul 151-742 (Korea, Republic of)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer Experiment was conducted on CHF under rolling condition in vertical tube. Black-Right-Pointing-Pointer CHF loop was mounted on rolling device to achieve rolling conditions. Black-Right-Pointing-Pointer Trends of CHF ratio as mass flux and pressure were studied. Black-Right-Pointing-Pointer Trends of CHF ratio under rolling motion was suggested using hypothetical CHF mechanism. - Abstract: This paper presents the characteristics of the critical heat flux (CHF) for the boiling of R-134a in vertical tube under rolling motion in a marine reactor. It is important to predict CHF of marine reactor under rolling motion in order to consider the safety margin of the reactor. MArine Reactor Moving Simulator (MARMS) test was conducted to measure the CHF of R-134a flowing upward in a uniformly heated vertical tube under rolling motion. A CHF loop mounted on rolling equipment, which can periodically roll from side to side through rotating by motor and mechanical power transmission gear. The CHF tests were performed in a 9.5 mm I.D. test section with heated length of 1 m. Mass flux ranges from 285 kg/m{sup 2} s to 1300 kg/m{sup 2} s, inlet subcoolings from 3 to 38 Degree-Sign C and outlet pressures from 1.3 to 2.4 bar, respectively. Amplitudes of rolling range from 15 Degree-Sign to 40 Degree-Sign and period from 6 to 12 s. Fluid-to-fluid (FTF) scaling was applied to convert the test matrix of MARMS from water to R-134a equivalent conditions. CHF ratios (ratio of the CHF under rolling condition to the stationary CHF) as mass flux and pressure in rolling motion are quite different from those of other existing transient CHF experiments. For the mass fluxes below 500 kg/m{sup 2} s (region of relative low mass flux) at 13, 16 bar, CHF ratios seem smaller than unit but in region (region of relative high mass flux) where mass fluxes are above 500 kg/m{sup 2} s, it was found that the ratios increased. Moreover, rolling CHFs tend to enhance

  4. Contact Angle Effects in Boiling Heat Transfer

    OpenAIRE

    Urquiola, Erwin; Fujita, Yasunobu

    2002-01-01

    This paper reports boiling experiments with pure water and surfactant solutions of SDS on horizontal heating surface. The static contact angle, rather than the surface tension value, was found to be the leading factor for the results and probably its prev

  5. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  6. The Neutrino Mass Hierarchy at Reactor Experiments now that theta13 is Large

    CERN Document Server

    Ciuffoli, Emilio; Zhang, Xinmin

    2012-01-01

    Now that theta13 is known to be large, a medium baseline reactor experiment can observe the fine structure of the electron antineutrino survival probability curve, approximately periodic oscillations in L/E with wavelength 4pi/Delta M^2_31. The periodicity with respect to L/E is broken by 2-3 oscillations which, in the case of the normal (inverted) hierarchy, shift the first 16 oscillations nearly 1% higher (lower) and move the next 16 lower (higher). The energy of each peak determines a particular combination of the mass differences, for example cos^2(theta12)Delta M^2_31 + sin^2(theta12)Delta M^2_32 for all peaks visible at baselines under 40 km. Comparing these combinations with each other or with NOvA results one can in principle determine the mass hierarchy. Alternately, as the Fourier transforms of the 1-3 and 2-3 oscillation probabilities are out of phase by the 1-2 oscillation probability, near the maximum of the 1-2 oscillation the complex phase of the total survival probability can be used to determ...

  7. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    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.

  8. Skylab experiments. Volume 4: Life sciences

    Science.gov (United States)

    1973-01-01

    The life sciences experiments conducted during Skylab missions are discussed. The general categories of the experiments are as follows: (1) mineral and hormonal balance, (2) hematology and immunology, (3) cardiovascular status, (4) energy expenditure, (5) neurophysiology, and (7) biology. Each experiment within the general category is further identified with respect to the scientific objectives, equipment used, performance, and data to be obtained.

  9. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John (Jack) K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  10. Constraints on very light sterile neutrinos from \\theta_{13}-sensitive reactor experiments

    CERN Document Server

    Palazzo, Antonio

    2013-01-01

    Three dedicated reactor experiments, Double Chooz, RENO and Daya Bay, have recently performed a precision measurement of the third standard mixing angle \\theta_{13} exploiting a multiple baseline comparison of nu_e -> nu_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\

  11. In-Reactor Oxidation of Zircaloy-4 Under Low Water Vapor Pressures

    Energy Technology Data Exchange (ETDEWEB)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330° and 370°C). Data from these tests will be used to support fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex- reactor test results were performed to evaluate the influence of irradiation.

  12. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    Energy Technology Data Exchange (ETDEWEB)

    Luscher, Walter G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Senor, David J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clayton, Kevin K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Longhurst, Glen R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ºC). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  13. Geysering in boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Aritomi, Masanori; Takemoto, Takatoshi [Tokyo Institute of Technology, Tokyo (Japan); Chiang, Jing-Hsien [Japan NUS Corp. Ltd., Toyko (Japan)] [and others

    1995-09-01

    A concept of natural circulation BWRs such as the SBWR has been proposed and seems to be promising in that the primary cooling system can be simplified. The authors have been investigating thermo-hydraulic instabilities which may appear during the start-up in natural circulation BWRs. In our previous works, geysering was investigated in parallel boiling channels for both natural and forced circulations, and its driving mechanism and the effect of system pressure on geysering occurrence were made clear. In this paper, geysering is investigated in a vertical column and a U-shaped vertical column heated in the lower parts. It is clarified from the results that the occurrence mechanism of geysering and the dependence of system pressure on geysering occurrence coincide between parallel boiling channels in circulation systems and vertical columns in non-circulation systems.

  14. Pool boiling on rectangular fins with tunnel-pore structure

    Directory of Open Access Journals (Sweden)

    Pastuszko A.

    2013-04-01

    Full Text Available Complex experimental investigations were conducted in the area of pool boiling heat transfer on extended surfaces with internal tunnels limited by perforated foil. The experiments were carried out for water and R-123 at atmospheric pressure. The tunnel surfaces were fabricated from 0.05 – 0.1 mm thick perforated copper foil (pore diameters: 0.3, 0.4, 0.5 mm sintered with mini-fins formed by 5 and 10 mm high rectangular fins and horizontal inter-fin surface. The effect of the main fin height, pore diameters and tunnel pitch on nucleate pool boiling was examined. Substantial enhancement of heat transfer coefficient was observed for the investigated surfaces. The highest increase in the heat transfer coefficient was obtained for the 10 mm high fins – about 50kW/m2K for water and 15 kW/m2K for R-123. The investigated surfaces showed boiling heat transfer coefficients similar to those of existing tunnel-pore structures.

  15. A Review of Boiling Heat Transfer Processes at High Heat Flux

    Science.gov (United States)

    1991-04-01

    liquid metals) which can lead to explosive boiling (known as bumping) that can lead to structural damage to hardware. 3 Transition boiling occurs between...to initiate boiling, in some cases having an explosive transition that can cause structural damage to hardware. A thorough understanding of boiling...graphical correlations for the pressure drops encountered in their experiments. About the same time, Staub and Walmet (Ref. 173) identified the two regions

  16. Oscillate boiling from microheaters

    Science.gov (United States)

    Li, Fenfang; Gonzalez-Avila, S. Roberto; Nguyen, Dang Minh; Ohl, Claus-Dieter

    2017-01-01

    We report about an intriguing boiling regime occurring for small heaters embedded on the boundary in subcooled water. The microheater is realized by focusing a continuous wave laser beam to about 10 μ m in diameter onto a 165-nm-thick layer of gold, which is submerged in water. After an initial vaporous explosion a single bubble oscillates continuously and repeatedly at several 100 kHz albeit with constant laser power input. The microbubble's oscillations are accompanied with bubble pinch-off, leading to a stream of gaseous bubbles in the subcooled water. The self-driven bubble oscillation is explained with a thermally kicked oscillator caused by surface attachment and by the nonspherical collapses. Additionally, Marangoni stresses induce a recirculating streaming flow which transports cold liquid towards the microheater, reducing diffusion of heat along the substrate and therefore stabilizing the phenomenon to many million cycles. We speculate that this oscillate boiling regime may overcome the heat transfer thresholds observed during the nucleate boiling crisis and offers a new pathway for heat transfer under microgravity conditions.

  17. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  18. S∧4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

    Science.gov (United States)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The S∧4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at ~1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S∧4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 ¢/K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S∧4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 ¢/K -0.9073 and 0.8502 $/atom%) are the lowest.

  19. Heat transfer correlation development and assessment: a summary and assessment of return to nucleate boiling phenomena during blowdown tests conducted at the Idaho National Engineering Laboratory (INEL). [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Eaton, A. M.; Tolman, E. L.

    1979-04-01

    The data are presented which were obtained in Loss-of-Coolant Experiments (LOCE) at Idaho National Engineering Laboratory (INEL) which demonstrate the presence of cladding rewetting after the critical heat flux has been exceeded as a viable cooling mechanism during the blowdown phase of a LOCE. A brief review of the mechanisms associated with the boiling crisis and rewetting is also provided. The relevance of INEL LOCE rewetting data to nuclear reactor licensing Evaluation Model Requirements is considered, and the conclusion is made that the elimination of rewetting and return to nucleate boiling (RNB) in Evaluation Models represents a definite conservatism.

  20. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  1. Waterproofed photomultiplier tube assemblies for the Daya Bay reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chow, Ken [Engineering Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Cummings, John [Department of Physics and Astronomy, Siena College, Loudonville, NY 12211 (United States); Edwards, Emily [Engineering Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Edwards, William [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Ely, Ry [Department of Physics, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Hoff, Matthew [Engineering Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Lebanowski, Logan [Department of Physics, University of Houston, Houston, TX 77204-5005 (United States); Li, Bo; Li, Piyi [School of Physics, Shandong University, Jinan 250100 (China); Lin, Shih-Kai [Department of Physics, University of Houston, Houston, TX 77204-5005 (United States); Liu, Dawei [Department of Physics, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Liu, Jinchang [Key Laboratory of Particle Astrophysics, Institute of High Energy Physics, Beijing 100049 (China); Luk, Kam-Biu, E-mail: k_luk@berkeley.edu [Department of Physics, University of California, Berkeley, CA 94720 (United States); Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Miao, Jiayuan [School of Physics, Shandong University, Jinan 250100 (China); Napolitano, Jim [Department of Physics, Temple University, Philadelphia, PA 19122 (United States); Ochoa-Ricoux, Juan Pedro [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Peng, Jen-Chieh [Department of Physics, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Qi, Ming [Department of Physics, Nanjing University, Nanjing 210000 (China); and others

    2015-09-11

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

  2. Waterproofed Photomultiplier Tube Assemblies for the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Chow, Ken; Edwards, Emily; Edwards, William; Ely, Ry; Hoff, Matthew; Lebanowski, Logan; Li, Bo; Li, Piyi; Lin, Shih-Kai; Liu, Dawei; Liu, Jinchang; Luk, Kam-Biu; Miao, Jiayuan; Napolitano, Jim; Ochoa-Ricoux, Juan Pedro; Peng, Jen-Chieh; Qi, Ming; Steiner, Herbert; Stoler, Paul; Stuart, Mary; Wang, Lingyu; Yang, Changgen; Zhong, Weili

    2015-01-01

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

  3. Boiling incipience and convective boiling of neon and nitrogen

    Science.gov (United States)

    Papell, S. S.; Hendricks, R. C.

    1977-01-01

    Forced convection and subcooled boiling heat transfer data for liquid nitrogen and liquid neon were obtained in support of a design study for a 30 tesla cryomagnet cooled by forced convection of liquid neon. This design precludes nucleate boiling in the flow channels as they are too small to handle vapor flow. Consequently, it was necessary to determine boiling incipience under the operating conditions of the magnet system. The cryogen data obtained over a range of system pressures, fluid flow rates, and applied heat fluxes were used to develop correlations for predicting boiling incipience and convective boiling heat transfer coefficients in uniformly heated flow channels. The accuracy of the correlating equations was then evaluated. A technique was also developed to calculate the position of boiling incipience in a uniformly heated flow channel. Comparisons made with the experimental data showed a prediction accuracy of plus or minus 15 percent

  4. Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

    DEFF Research Database (Denmark)

    Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

    2013-01-01

    to reactor modelling. Before version 9.5, Geant4 HP thermal scattering model (i.e. the S(α; β) model ) supports only three bounded isotopes, namely, H in water and polyethylene, and C in graphite. Newly supported materials include D in heavy water, O and Be in beryllium oxide, H and Zr in zirconium hydride......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...

  5. Systematic impact of spent nuclear fuel on θ13 sensitivity at reactor neutrino experiment

    Institute of Scientific and Technical Information of China (English)

    AN Feng-Peng; TIAN Xin-Chun; ZHAN Liang; CAO Jun

    2009-01-01

    Reactor neutrino oscillation experiments, such as Daya Bay, Double Chooz and RENO are designed to determine the neutrino mixing angle θ13 with a sensitivity of 0.01--0.03 in sin2 2θ13 at 90% confidence level, an improvement over the current limit by more than one order of magnitude. The control of systematic uncertainties is critical to achieving the sin2 2θ13 sensitivity goal of these experiments. Antineutrinos emitted from spent nuclear fuel (SNF) would distort the soft part of energy spectrum and may introduce a non-negligible systematic uncertainty. In this article, a detailed calculation of SNF neutrinos is performed taking account of the operation of a typical reactor and the event rate in the detector is obtained. A further estimation shows that the event rate contribution of SNF neutrinos is less than 0.2% relative to the reactor neutrino signals. A global χ2 analysis shows that this uncertainty will degrade the θ13 sensitivity at a negligible level.

  6. Recent numerical simulations and experiments on coolability of debris beds during severe accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J., E-mail: joerg.starflinger@ike.uni-stuttgart.de; Buck, M.; Hartmann, A.; Kulenovic, R.; Leininger, S.; Rahman, S.; Rashid, M.

    2015-12-01

    Highlights: • Investigation on coolability of three-dimensional debris beds has been performed. • Computer code MEWA (Melt Water) is introduced and described briefly. • Validation experiments have been carried out in DEBRIS facility. • Comparison of MEWA simulations and DEBRIS experiments show good agreement. • Example simulation on reactor scale was performed to explain the analysis method. - Abstract: In the course of a severe accident in light water reactors with core degradation, so-called debris beds can be formed inside the reactor pressure vessel or in the reactor cavity. The strategy to analyse the coolability of such debris beds with both experiments and numerical simulations is discussed. The numerical simulations are carried out with MEWA (MElt WAter) code, being developed at the institute for the prediction of the thermal-hydraulic conditions inside a debris bed, including the prediction of dryout heat flux. The simulations show good agreement with experimental data of the DEBRIS experiments.

  7. New Revelation of Lightning Ball Observation and Proposal for a Nuclear Reactor Fusion Experiment

    CERN Document Server

    Tar, Domokos

    2009-01-01

    In this paper, the author brings further details regarding his Lightning Ball observation that were not mentioned in the first one (Ref.1-2). Additionally, he goes more into detail as the three forces that are necessary to allow the residual crescent form the hydrodynamic vortex ring to shrink into a sphere.Further topics are the similarities and analogies between the Lightning Ball formation's theory and the presently undertaken Tokamak-Stellarator-Spheromak fusion reactor experiments. A new theory and its experimental realisation are proposed as to make the shrinking of the hot plasma of reactors into a ball possible by means of the so called long range electromagnetic forces. In this way,the fusion ignition temperature could possibly atteined.

  8. Experiments and simulations of gas-solid flow in an airlift loop reactor

    Institute of Scientific and Technical Information of China (English)

    Chaoyu Yan; Chunxi Lu; Yiping Fan; Rui Cao; Yansheng Liu

    2011-01-01

    The hydrodynamics in a gas-solid airlift loop reactor was investigated systematically using experimental measurements and CFD simulation. In the experiments, the time averaged parameters, such as solid fraction and particle velocity, were measured by optical fiber probe. In the simulation, the modified Gidaspow drag model accounting for the interparticles clustering was incorporated into the Eulerian-Eulerian CFD model with particulate-phase kinetic theory. Predicted values of solid fraction and particle velocity were compared with experimental results, validating the drag model and the simulation. The results show that the profiles of particle velocity and solid fraction are uniform in annulus. However, the core-annulus structure appears in other three regions (draft tube region, bottom region and particle diffiuence region),which presents the similar heterogeneous feature of aggregative fiuidization usually occurred in normal fiuidized beds. Simulated profiles of panicle residence time distribution indicate that the airlift loop reactor should be characterized by near perfect mixing.

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

    Directory of Open Access Journals (Sweden)

    Yu-Feng Li

    2014-11-01

    Full Text Available We discuss reactor antineutrino oscillations with non-standard interactions (NSIs at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters.

  10. Boiling flow through diverging microchannel

    Indian Academy of Sciences (India)

    V S Duryodhan; S G Singh; Amit Agrawal

    2013-12-01

    An experimental study of flow boiling through diverging microchannel has been carried out in this work, with the aim of understanding boiling in nonuniform cross-section microchannel. Diverging microchannel of 4° of divergence angle and 146 m hydraulic diameter (calculated at mid-length) has been employed for the present study with deionised water as working fluid. Effect of mass flux (118–1182 kg/m2-s) and heat flux (1.6–19.2 W/cm2) on single and two-phase pressure drop and average heat transfer coefficient has been studied. Concurrently, flow visualization is carried out to document the various flow regimes and to correlate the pressure drop and average heat transfer coefficient to the underlying flow regime. Four flow regimes have been identified from the measurements: bubbly, slug, slug–annular and periodic dry-out/rewetting. Variation of pressure drop with heat flux shows one maxima which corresponds to transition from bubbly to slug flow. It is shown that significantly large heat transfer coefficient (up to 107 kW/m2-K) can be attained for such systems, for small pressure drop penalty and with good flow stability.

  11. Experimental investigation on the boiling heat transfer of nanofluids on a flat plate in the presence of a magnetic field

    Science.gov (United States)

    Abdollahi, Ali; Reza Salimpour, Mohammad

    2016-11-01

    In this paper, the pool boiling heat transfer of Fe3O4 -deionized (DI) water as a magnetic nanofluid has been experimentally analyzed in the atmospheric pressure. The applied nanofluid within this research has been synthesized through a single step to retain a high stability. The repeatability and precision of the testing device with deionized water show a good agreement with the equations introduced in previous studies. Parametric studies on magnetic field, surface roughness, and magnetic nanofluid concentration are performed to reveal various aspects of the boiling heat transfer. In order to study the surface roughness, two surfaces with high average roughness (480nm) and low average roughness (7.3nm) were used. The obtained results indicate that the boiling heat transfer on the rough surface increases when raising the nanofluid concentration up to 0.1% volume concentration. In addition, it is observed that there is an optimum 0.1% volume concentration for the nanofluid which makes the boiling heat transfer coefficient increase up to 43%. Moreover, the heat transfer of a nanofluid with volume concentration of 0.1% is greater for the rough surface compared with the smooth one. The results of the experiments indicate that adding nanoparticles would not necessarily increase the boiling heat transfer coefficient. In fact, the surface roughness and the magnetic field gradient on the boiling surface were the main factors that could affect the boiling heat transfer coefficient significantly. The simultaneous analysis of magnetic field, surface roughness, and nanofluid concentration reveals that the boiling heat transfer coefficient of the magnetic nanofluid with 0.1% volume concentration in the presence of a magnetic field on the rough surface is higher than on the smooth surface. Our findings show that this increase is associated to the increase of nucleation sites concentration and bubble formation sites for the rough surface.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  13. Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

    2014-10-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

  14. Direct Numerical Simulation and Visualization of Subcooled Pool Boiling

    Directory of Open Access Journals (Sweden)

    Tomoaki Kunugi

    2014-01-01

    Full Text Available A direct numerical simulation of the boiling phenomena is one of the promising approaches in order to clarify their heat transfer characteristics and discuss the mechanism. During these decades, many DNS procedures have been developed according to the recent high performance computers and computational technologies. In this paper, the state of the art of direct numerical simulation of the pool boiling phenomena during mostly two decades is briefly summarized at first, and then the nonempirical boiling and condensation model proposed by the authors is introduced into the MARS (MultiInterface Advection and Reconstruction Solver developed by the authors. On the other hand, in order to clarify the boiling bubble behaviors under the subcooled conditions, the subcooled pool boiling experiments are also performed by using a high speed and high spatial resolution camera with a highly magnified telescope. Resulting from the numerical simulations of the subcooled pool boiling phenomena, the numerical results obtained by the MARS are validated by being compared to the experimental ones and the existing analytical solutions. The numerical results regarding the time evolution of the boiling bubble departure process under the subcooled conditions show a very good agreement with the experimental results. In conclusion, it can be said that the proposed nonempirical boiling and condensation model combined with the MARS has been validated.

  15. Construction of elements data base of JRR-3, 4 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Akira; Iwata, Yoshihiro [Akita Univ. (Japan); Sawahata, Hiroyuki

    1998-01-01

    Two kinds of the basic data base were constructed. One of them contains stable isotope information (atomic number, mass number, atomic weight and isotope ratio) and other consists of radionuclide information (atomic number, mass number, half-life, {gamma}-ray average energy, {beta}-ray average energy and {beta}+ray release rate). We tried to add the spectrum data determined by experiments, but many kinds of radionuclide in the sample could not separated. (S.Y.)

  16. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  17. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  18. Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Seidel, Tobias; Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Deendarlianto

    2011-09-15

    In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, a model of the hot leg of a pressurised water reactor was built at FZD. The hot leg model is operated in the pressure chamber of the TOPFLOW test facility, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the chamber. This technique makes it possible to visualise the two-phase flow through large windows, also at reactor-typical pressure levels. In order to optimise the optical observation possibilities, the test section was designed with a rectangular cross-section. Experiments were performed with air and water at 1.5 and 3.0 bar at room temperature as well as with steam and water at 15, 30 and 50 bar and the corresponding saturation temperature (i.e. up to 264 C). The total of 194 runs are divided into 4 types of experiments covering stationary co-current flow, counter-current flow, flow without water circulation and transient counter-current flow limitation (CCFL) experiments. This report provides a detailed documentation of the experiments including information on the experimental setup, experimental procedure, test matrix and on the calibration of the measuring devices. The available data is described and data sheets were arranged for each experiment in order to give an overview of the most important parameters. For the cocurrent flow experiments, water level histograms were arranged and used to characterise the flow in the hot leg. In fact, the form of the probability distribution was found to be sensitive to the boundary conditions and, therefore, is useful for the CFD comparison. Furthermore, the flooding characteristics of the hot leg model plotted in terms of the classical Wallis parameter or Kutateladze number were found to fail to properly correlate the data of the air/water and steam/water series. Therefore, a modified Wallis parameter is proposed, which

  19. Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kulesza, J.A.; Fero, A.H. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Rouden, J.; Green, E.L. [Vattenfall/Ringhals AB, 432 85 Vaeroebacka (Sweden)

    2011-07-01

    A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the {sup 93}Nb (n,n') {sup 93m}Nb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  1. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  2. Mass Hierarchy Resolution in Reactor Anti-neutrino Experiments: Parameter Degeneracies and Detector Energy Response

    Energy Technology Data Exchange (ETDEWEB)

    X. Qian, D. A. Dwyer, R. D. McKeown, P. Vogel, W. Wang, C. Zhang`

    2013-02-01

    Determination of the neutrino mass hierarchy using a reactor neutrino experiment at ∼60  km is analyzed. Such a measurement is challenging due to the finite detector resolution, the absolute energy scale calibration, and the degeneracies caused by current experimental uncertainty of |Δm{sub 32}{sup 2}|. The standard {chi}{sup 2} method is compared with a proposed Fourier transformation method. In addition, we show that for such a measurement to succeed, one must understand the nonlinearity of the detector energy scale at the level of a few tenths of percent.

  3. First Test of Lorentz Violation with a Reactor-based Antineutrino Experiment

    CERN Document Server

    Abe, Y; Anjos, J C dos; 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; Erickson, A; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Fischer, V; 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; Habib, S; Hagner, C; Hara, T; Hartmann, F X; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G A; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L N; Kamyshkov, Y; Kaplan, D M; Katori, T; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; 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; Meyer, M; 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; Pronost, G; Reichenbacher, J; Reinhold, B; Remoto, A; Röhling, M; Roncin, R; Roth, S; Rybolt, B; Sakamoto, Y; Santorelli, R; Sato, F; Schönert, S; Schoppmann, S; Schwetz, T; Shaevitz, M H; 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; Terao, K; Tonazzo, A; Toups, M; Thi, H H Trinh; Valdiviesso, G; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yanovitch, E; Yermia, F; Zimmer, V

    2012-01-01

    We present a search for Lorentz violation with 8249 candidate electron antineutrino events taken by the Double Chooz experiment in 227.9 live days of running. This analysis, featuring a search for a sidereal time dependence of the events, is the first test of Lorentz invariance using a reactor-based antineutrino source. No sidereal variation is present in the data and the disappearance results are consistent with sidereal time independent oscillations. Under the Standard-Model Extension (SME), we set the first limits on fourteen Lorentz violating coefficients associated with transitions between electron and tau flavor, and set two competitive limits associated with transitions between electron and muon flavor.

  4. Production of Gadolinium-loaded Liquid Scintillator for the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Beriguete, Wanda; Ding, Yayun; Hans, Sunej; Heeger, Karsten M; Hu, Liangming; Huang, Aizhong; Luk, Kam-Biu; Nemchenok, Igor; Qi, Ming; Rosero, Richard; Sun, Hansheng; Wang, Ruiguang; Wang, Yifang; Wen, Liangjian; Yang, Yi; Yeh, Minfang; Zhang, Zhiyong; Zhou, Li

    2014-01-01

    We report on the production and characterization of liquid scintillators for the detection of electron antineutrinos by the Daya Bay Reactor Neutrino Experiment. One hundred eighty-five tons of gadolinium-loaded (0.1% by mass) liquid scintillator (Gd-LS) and two hundred tons of unloaded liquid scintillator (LS) were successfully produced from a linear-alkylbenzene (LAB) solvent in six months. The scintillator properties, the production and purification systems, and the quality assurance and control (QA/QC) procedures are described.

  5. Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; Hans D. Gougar; John M. Ryskamp

    2013-09-01

    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. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government 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. This work can be viewed as an update to a project from the 1990’s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the “Cylindrical” variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a

  6. Subcooled boiling of nano-particle suspensions on Pt wires

    Institute of Scientific and Technical Information of China (English)

    LI Chunhui; WANG Buxuan; PENG Xiaofeng

    2004-01-01

    An experimental investigation is conducted to explore the subcooled boiling characteristics of nano-particle suspensions on Pt wires. Some phenomena are observed for the boiling of water-SiO2 nano-particle suspensions on Pt wires. The experiments show that there exist not any evident differences for boiling of pure water and of nano-particle suspensions at high heat fluxes. However, bubble overlap phenomenon can be easily found for nano-particle suspensions at low heat fluxes, which probably results from the increase of the attracter force between bubbles and of the bubble mass.

  7. Benchmark Data Through The International Reactor Physics Experiment Evaluation Project (IRPHEP)

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Dr. Enrico Sartori

    2005-09-01

    The International Reactor Physics Experiments Evaluation Project (IRPhEP) was initiated by the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency’s (NEA) Nuclear Science Committee (NSC) in June of 2002. The IRPhEP focus is on the derivation of internationally peer reviewed benchmark models for several types of integral measurements, in addition to the critical configuration. While the benchmarks produced by the IRPhEP are of primary interest to the Reactor Physics Community, many of the benchmarks can be of significant value to the Criticality Safety and Nuclear Data Communities. Benchmarks that support the Next Generation Nuclear Plant (NGNP), for example, also support fuel manufacture, handling, transportation, and storage activities and could challenge current analytical methods. The IRPhEP is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and is closely coordinated with the ICSBEP. This paper highlights the benchmarks that are currently being prepared by the IRPhEP that are also of interest to the Criticality Safety Community. The different types of measurements and associated benchmarks that can be expected in the first publication and beyond are described. The protocol for inclusion of IRPhEP benchmarks as ICSBEP benchmarks and for inclusion of ICSBEP benchmarks as IRPhEP benchmarks is detailed. The format for IRPhEP benchmark evaluations is described as an extension of the ICSBEP format. Benchmarks produced by the IRPhEP add new dimension to criticality safety benchmarking efforts and expand the collection of available integral benchmarks for nuclear data testing. The first publication of the "International Handbook of Evaluated Reactor Physics Benchmark Experiments" is scheduled for January of 2006.

  8. Pd-Ag Membrane Coupled to a Two-Zone Fluidized Bed Reactor (TZFBR for Propane Dehydrogenation on a Pt-Sn/MgAl2O4 Catalyst

    Directory of Open Access Journals (Sweden)

    Miguel Menéndez

    2013-05-01

    Full Text Available Several reactor configurations have been tested for catalytic propane dehydrogenation employing Pt-Sn/MgAl2O4 as a catalyst. Pd-Ag alloy membranes coupled to the multifunctional Two-Zone Fluidized Bed Reactor (TZFBR provide an improvement in propane conversion by hydrogen removal from the reaction bed through the inorganic membrane in addition to in situ catalyst regeneration. Twofold process intensification is thereby achieved when compared to the use of traditional fluidized bed reactors (FBR, where coke formation and thermodynamic equilibrium represent important process limitations. Experiments were carried out at 500–575 °C and with catalyst mass to molar flow of fed propane ratios between 15.1 and 35.2 g min mmol−1, employing three different reactor configurations: FBR, TZFBR and TZFBR + Membrane (TZFBR + MB. The results in the FBR showed catalyst deactivation, which was faster at high temperatures. In contrast, by employing the TZFBR with the optimum regenerative agent flow (diluted oxygen, the process activity was sustained throughout the time on stream. The TZFBR + MB showed promising results in catalytic propane dehydrogenation, displacing the reaction towards higher propylene production and giving the best results among the different reactor configurations studied. Furthermore, the results obtained in this study were better than those reported on conventional reactors.

  9. The KASKA project - a Japanese medium-baseline reactor-neutrino oscillation experiment to measure the mixing angle $\\theta_{13}$ -

    CERN Document Server

    Kuze, M

    2005-01-01

    A new reactor-neutrino oscillation experiment, KASKA, is proposed to measure the unknown neutrino-mixing angle $\\theta_{13}$ using the world's most powerful Kashiwazaki-Kariwa nuclear power station. It will measure a very small deficit of reactor-neutrino flux using three identical detectors, two placed just close to the sources and one at a distance of about 1.8km. Its conceptual design and physics reach are discussed.

  10. Neutron Spectral Brightness of Cold Guide 4 at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Winn, B L; Robertson, J L; Iverson, E B; Selby, D L, E-mail: winnbl@ornl.gov

    2010-11-01

    The High Flux Isotope Reactor resumed operation in June of 2007 with a supercritical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source with a reasonable flux at wavelengths greater than 4 A to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

  11. Neutron Spectral Brightness of Cold Guide 4 at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Winn, B. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Neutron Scattering Group; Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Neutron Facilities Development Div.; Robertson, J. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Neutron Facilities Development Div.; Iverson, Erik B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Neutron Facilities Development Div.; Selby, D. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Neutron Facilities Development Div.

    2009-05-03

    The High Flux Isotope Reactor resumed operation in June of 2007 with a super-critical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source at reasonable flux at wavelengths greater than 4 Å to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

  12. Neutron Spectral Brightness of Cold Guide 4 at the High Flux Isotope Reactor

    Science.gov (United States)

    Winn, B. L.; Robertson, J. L.; Iverson, E. B.; Selby, D. L.

    2010-11-01

    The High Flux Isotope Reactor resumed operation in June of 2007 with a supercritical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source with a reasonable flux at wavelengths greater than 4 Å to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

  13. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics with those using the isotopics generated by the SCALE-4

  14. Crystallization fouling of finned tubes during pool boiling: effect of fin density

    Energy Technology Data Exchange (ETDEWEB)

    Esawy, M.; Malayeri, M.R. [University of Stuttgart, Institute for Thermodynamics and Thermal Engineering (ITW), Stuttgart (Germany); Mueller-Steinhagen, H. [University of Stuttgart, Institute for Thermodynamics and Thermal Engineering (ITW), Stuttgart (Germany); German Aerospace Centre (DLR), Institute of Technical Thermodynamics, Stuttgart (Germany)

    2010-11-15

    Bubble characteristics such as density, size, frequency and motion are key factors that contribute to the superiority of nucleate pool boiling over other modes of heat transfer. Nevertheless, if heat transfer occurs in an environment prone to fouling, the very same parameters may lead to accelerated deposit formation due to concentration effects beneath the growing bubbles. This has led to the widely accepted design recommendation to maintain the heat transfer surface temperature below the boiling point if fouling may occur, e.g., in seawater desalination. The present paper aims at investigating the formation of deposits on finned tubes during nucleate pool boiling of CaSO{sub 4} solutions. The test finned tubes are low finned tubes with fin densities of 19 and 26 fins/in. made from Cu-Ni. The fouling experiments were carried out at atmospheric pressure for different heat fluxes ranging from 100 to 300 kW/m{sup 2} and a CaSO{sub 4} concentration of 1.6 g/L. For the sake of comparison, similar runs were performed with smooth stainless steel tubes. The results show that: (1) the fouling resistance decreases with increasing fin density, (2) fouling on the finned tubes was reduced with increasing nucleate boiling activity and (3) if any fouling layer occurred on the finned tubes it could be removed easily. (orig.)

  15. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Farnum, E.; Scharborough, K. [Los Alamos National Lab., NM (United States); Shikama, Tatsuo [and others

    1995-04-01

    The objective of this experiment is to determine the extent of degradation during neutron irradiation of electrical and optical properties of candidate dielectric materials. The goals are to identify promising dielectrics for ITER and other fusion machines for diagnostic applications and establish the basis for optimization of candidate materials. An experiment to measure radiation-induced electrical degradation (REID) in sapphire and MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260 {degree}C to a fluence of 3{times}10{sup 24} n/m{sup 2} (E>1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m.

  16. Fusion Reactor and Break-Even Experiment Based on Stabilized Liner Compression of Plasma

    Science.gov (United States)

    Turchi, Peter; Frese, Sherry; Frese, Michael

    2016-10-01

    An optimum regime, known as magnetized-target or magneto-inertial fusion (MTF/MIF), requires magnetic fields at megagauss levels, which are attainable by use of dynamic conductors called liners. The stabilized liner compressor (SLC) provides the basis for controlled implosion and re-capture of the liner for reversible energy exchange between liner kinetic energy and the internal energy of a magnetized-plasma target. This exchange requires rotational stabilization of Rayleigh-Taylor modes on the inner surface of the liner and pneumatically driven free-pistons that eliminate such modes at the outer surface. We discuss the implications of the SLC approach for the power reactor, a breakeven experiment, and intermediate experiments to develop the plasma target. Features include the importance of pneumatic drive and the liner-blanket for economic feasibility of MTF/MIF. Supported by ARPA-E ALPHA Program.

  17. Limits on nu_e and anti-nu_e disappearance from Gallium and reactor experiments

    CERN Document Server

    Acero, Mario A; Laveder, Marco

    2007-01-01

    The disappearance of electron neutrinos observed in the Gallium radioactive source experiments is analyzed in the framework of two-neutrino mixing. It is shown that there is an indication of neutrino disappearance due to neutrino oscillations with sin^2 2 theta >~ 0.04 and Delta m^2 >~ 0.1 eV^2. The compatibility of this result with the data of the Bugey and Chooz reactor short-baseline antineutrino disappearance experiments is studied. It is found that the Bugey data present a weak indication in favor of neutrino oscillations with 0.02 <~ sin^2 2 theta <~ 0.08 and Delta m^2 =~ 1.85 eV^2, which is compatible with the Gallium allowed region of the mixing parameters. This indication persists in the combined analyses of Bugey and Chooz data, of Gallium and Bugey data, and of Gallium, Bugey, and Chooz data.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

  19. Transition boiling heat transfer and the film transition regime

    Science.gov (United States)

    Ramilison, J. M.; Lienhard, J. H.

    1987-01-01

    The Berenson (1960) flat-plate transition-boiling experiment has been recreated with a reduced thermal resistance in the heater, and an improved access to those portions of the transition boiling regime that have a steep negative slope. Tests have been made in Freon-113, acetone, benzene, and n-pentane boiling on horizontal flat copper heaters that have been mirror-polished, 'roughened', or teflon-coated. The resulting data reproduce and clarify certain features observed by Berenson: the modest surface finish dependence of boiling burnout, and the influence of surface chemistry on both the minimum heat flux and the mode of transition boiling, for example. A rational scheme of correlation yields a prediction of the heat flux in what Witte and Lienhard (1982) previously identified as the 'film-transition boiling' region. It is also shown how to calculate the heat flux at the boundary between the pure-film, and the film-transition, boiling regimes, as a function of the advancing contact angle.

  20. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  1. SAS wave experiment on board Magion 4

    Directory of Open Access Journals (Sweden)

    J. Błęcki

    Full Text Available A short description of the SAS (subsatellite analyser of spectra wave experiment on board the Magion-4 subsatellite is given. We present first measurements of the magnetic-field fluctuations in the frequency range 32–2000 Hz obtained in the magnetotail during the disturbed period at the magnetopause and in the polar cusp.

  2. Report on Experiment AD-4/ACE

    CERN Document Server

    Holzscheiter, M H

    2014-01-01

    After 7 years running the AD-4 Experiment at antiproton momentum of 502 MeV/c we have now assembled a complete data set and are preparing to combine all years into a single analysis of RBE vs. Depth for an antiproton beam stopping in water. We describe experimental challenges and computational issues which need to be resolved to achieve this final step.

  3. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  4. Preliminary Study on LiF4-ThF4-PuF4 Utilization as Fuel Salt of miniFUJI Molten Salt Reactor

    Science.gov (United States)

    Waris, Abdul; Aji, Indarta K.; Pramuditya, Syeilendra; Widayani; Irwanto, Dwi

    2016-08-01

    miniFUJI reactor is molten salt reactor (MSR) which is one type of the Generation IV nuclear energy systems. The original miniFUJI reactor design uses LiF-BeF2-ThF4-233UF4 as a fuel salt. In the present study, the use of LiF4-ThF4-PuF4 as fuel salt instead of LiF-BeF2-ThF4-UF4 will be discussed. The neutronics cell calculation has been performed by using PIJ (collision probability method code) routine of SRAC 2006 code, with the nuclear data library is JENDL-4.0. The results reveal that the reactor can attain the criticality condition with the plutonium concentration in the fuel salt is equal to 9.16% or more. The conversion ratio diminishes with the enlarging of plutonium concentration in the fuel. The neutron spectrum of miniFUJI MSR with plutonium fuel becomes harder compared to that of the 233U fuel.

  5. Experiences on Grid Production for GEANT4

    CERN Document Server

    Lamanna, M; Méndez-Lorenzo, P; Moscicki, J; Ribon, A

    2007-01-01

    Geant4 is a worldwide collaboration of scientists and software engineers whose goal is to develop, maintain and provide support for the Geant4 package, representing a general purpose Monte Carlo toolkit for simulating the propagation of particles through matter. Geant4 is currently employed in a row of particle physics experiments (BaBar, HARP, ATLAS, CMS, LHCb), and is also applied in other areas like space science, medical applications, and radiation studies. Due to the complexity of the Geant4 code and to the broad spectrum of possible configurations - involving a variety of physical processes for different source particles, particle energies and target or shielding materials - an intensive testing of new release candidates is mandatory to thoroughly test each of its components, especially before major releases, generally twice a year. Regression tests are required to be performed by the Geant4 team within a short period of time (2 to 3 weeks), basically demanding vast computational resources (equivalent t...

  6. Film boiling of mercury droplets

    Science.gov (United States)

    Baumeister, K. J.; Schoessow, G. J.; Chmielewski, C. E.

    1975-01-01

    Vaporization times of mercury droplets in Leidenfrost film boiling on a flat horizontal plate are measured in an air atmosphere. Extreme care was used to prevent large amplitude droplet vibrations and surface wetting; therefore, these data can be compared to film boiling theory. For these data, diffusion from the upper surface of the drop is a dominant mode of mass transfer from the drop. A closed-form analytical film boiling theory is developed to account for the diffusive evaporation. Reasonable agreement between data and theory is seen.

  7. Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, S.; Ponsard, B.; Koonen, E.

    2007-07-15

    The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

  8. Electrical Capacitance Volume Tomography for the Packed Bed Reactor ISS Flight Experiment

    Science.gov (United States)

    Marashdeh, Qussai; Motil, Brian; Wang, Aining; Liang-Shih, Fan

    2013-01-01

    Fixed packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a highly desirable unit operation for long duration life support systems in space. 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. To validate these models, the instantaneous distribution of the gas and liquid phases must be measured.Electrical Capacitance Volume Tomography (ECVT) is a non-invasive imaging technology recently developed for multi-phase flow applications. It is based on distributing flexible capacitance plates on the peripheral of a flow column and collecting real-time measurements of inter-electrode capacitances. Capacitance measurements here are directly related to dielectric constant distribution, a physical property that is also related to material distribution in the imaging domain. Reconstruction algorithms are employed to map volume images of dielectric distribution in the imaging domain, which is in turn related to phase distribution. ECVT is suitable for imaging interacting materials of different dielectric constants, typical in multi-phase flow systems. ECVT is being used extensively for measuring flow variables in various gas-liquid and gas-solid flow systems. Recent application of ECVT include flows in risers and exit regions of circulating fluidized beds, gas-liquid and gas-solid bubble columns, trickle beds, and slurry bubble columns. ECVT is also used to validate flow models and CFD simulations. The technology is uniquely qualified for imaging phase concentrations in packed bed reactors for the ISS flight experiments as it exhibits favorable features of compact size, low profile sensors, high imaging speed, and

  9. Characteristics of Al Alloy as a Material for Hydrolysis Reactor of NaBH{sub 4}

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hyeon-Seong; Oh, Sung-June; Jeong, Jae-Jin; Na, Il-Chai; Chu, Cheun-Ho; Park, Kwon-Pil [Sunchon National University, Suncheon (Korea, Republic of); Chu, Cheun-Ho [ETIS Co, Gimpo (Korea, Republic of)

    2015-12-15

    Aluminum alloy was examined as a material of low weight reactor for hydrolysis of NaBH{sub 4}. Aluminum is dissolved with alkali, but there is NaOH as a stabilizer in NaBH{sub 4} solution. To decrease corrosion rate of aluminum, decrease NaOH concentration and this result in loss of NaBH{sub 4} during storage of NaBH{sub 4} solution. Therefore stability of NaBH{sub 4} and corrosion of aluminum should be considered in determining the optimum NaOH concentration. NaBH{sub 4} stability and corrosion rate of aluminum were measured by hydrogen evolution rate. NaBH{sub 4} stability was tested at 20-50 .deg. C and aluminum corrosion was measured at 60-90 .deg. C. The optimum concentration of NaOH was 0.3 wt%, considering both NaBH{sub 4} stability and aluminun corrosion. NaBH{sub 4} hydrolysis reaction continued 200min in aluminum No 6061 alloy reactor with 0.3 wt% NaOH at 80-90 .deg. C.

  10. Boiling Heat Transfer in Circulating Fluidized Beds

    Institute of Scientific and Technical Information of China (English)

    张利斌; 李修伦

    2001-01-01

    A model is proposed to predict boiling heat transfer coefficient in a three-phase circulating fluidized bed (CFB), which is a new type of evaporation boiling means for enhancing heat transfer and preventing fouling. To verify the model, experiments are conducted in a stainless steel column with 39 mm ID and 2.0 m height, in which the heat transfer coefficient is measured for different superficial velocities, steam pressures, particle concentrations and materials of particle. As the steam pressure and particle concentrations increase, the heat transfer coefficient in the bed increases. The heat transfer coefficient increases with the liquid velocity but it exhibits a local minimum.The heat transfer coefficient is correlated with cluster renewed model and two-mechanism method. The prediction of the model is in good agreement with experimental data.

  11. Boiling Heat Transfer in Circulating Fluidized Beds

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A model is proposed to predict boiling heat transfer coefficient in a three-phase circulating fluidized bed (CFB), which is a new type of evaporation boiling means for enhancing heat transfer and preventing fouling. To verify the model, experiments are conducted in a stainless steel column with 39mm ID and 2.0m height, in which the heat transfer coefficient is measured for different superficial velocities, steam pressures, particle concentrations and materials of particle. As the steam pressure and particle concentrations increase, the heat transfer coefficient in the bed increases. The heat transfer coefficient increases with the liquid velocity but it exhibits a local minimum. The heat transfer coefficient is correlated with cluster renewed model and two-mechanism method. The prediction of the model is in good agreement with experimental data.

  12. Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

    Energy Technology Data Exchange (ETDEWEB)

    Paniagua, J.; Rohatgi, U.S.; Prasad, V.

    1996-10-01

    RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).

  13. Flow boiling test of GDP replacement coolants

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.H. [comp.

    1995-08-01

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

  14. Boiling local heat transfer enhancement in minichannels using nanofluids.

    Science.gov (United States)

    Chehade, Ali Ahmad; Gualous, Hasna Louahlia; Le Masson, Stephane; Fardoun, Farouk; Besq, Anthony

    2013-03-18

    This paper reports an experimental study on nanofluid convective boiling heat transfer in parallel rectangular minichannels of 800 μm hydraulic diameter. Experiments are conducted with pure water and silver nanoparticles suspended in water base fluid. Two small volume fractions of silver nanoparticles suspended in water are tested: 0.000237% and 0.000475%. The experimental results show that the local heat transfer coefficient, local heat flux, and local wall temperature are affected by silver nanoparticle concentration in water base fluid. In addition, different correlations established for boiling flow heat transfer in minichannels or macrochannels are evaluated. It is found that the correlation of Kandlikar and Balasubramanian is the closest to the water boiling heat transfer results. The boiling local heat transfer enhancement by adding silver nanoparticles in base fluid is not uniform along the channel flow. Better performances and highest effect of nanoparticle concentration on the heat transfer are obtained at the minichannels entrance.

  15. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized- water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SAS2H analytical sequence in SCALE-4. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code sequence was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE-4 criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for analysis of each critical configuration. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power

  16. Measurement of Nucleate Pool Boiling Heat Transfer Limit using Fuel Cladding Material

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chi Young; Shin, Chang Hwan; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    Zircaloy has been widely used as a fuel cladding material of light water reactor for more than three decades because it has a lower neutron absorption cross section and cracking rate. Recently, HANA-6 has been developed in KAERI (Korea Atomic Energy Research Institute) as the advanced fuel cladding for high burn-up fuel. Generally, under the normal and accident operating conditions of a nuclear reactor, the nuclear fuel cladding of zirconium based alloys undergoes the surface change, and the oxide layer can be formed. In such a case, the previous CHF correlations should be assessed and examined using the experimental results for not a fresh zircaloy surface but an oxidized one, to predict and examine the thermal margin and safety of a nuclear reactor core. Therefore, the experimental data using the oxidized zircaloy surface need to be provided quantitatively. In this paper, the CHF in saturated water pool boiling is measured and discussed using the specimens of zircaloy-4, HANA-6, and oxidized zircaloy-4 in high temperature air environment. The CHF of zircaloy-4, HANA-6, and oxidized surface was tested. Zircaloy-4 and HANA-6 had a similar CHF performance. This is because both are the zirconium based alloys, and appear the almost same water contact angle. On the other hands, the oxidized specimen became to be higher CHF than plain zircaloy-4 and HANA-6 specimens, due to smaller water contact angle (i. e., good hydrophilicity of specimen). The Kandlikar's (2001) correlation reasonably predicted the present experimental data.

  17. Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

    Science.gov (United States)

    Niederauer, G. F.

    1973-01-01

    The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

  18. GEANT4 Simulation of the NPDGamma Experiment

    Science.gov (United States)

    Frlez, Emil

    2014-03-01

    The n-> + p --> d + γ experiment, currently taking data at the Oak Ridge SNS facility, is a high-precision measurement of weak nuclear forces at low energies. Detecting the correlation between the cold neutron spin and photon direction in the capture of neutrons on Liquid Hydrogen (LH) target, the experiment is sensitive to the properties of neutral weak current. We have written a GEANT4 Monte Carlo simulation of the NPDGamma detector that, in addition to the active CsI detectors, also includes different targets and passive materials as well as the beam line elements. The neutron beam energy spectrum, its profiles, divergencies, and time-of-flight are simulated in detail. We have used the code to cross-calibrate the positions of (i) polarized LH target, (ii) Aluminum target, and (iii) CCl4 target. The responses of the 48 CsI detectors in the simulation were fixed using data taken on the LH target. Both neutron absorption as well as scattering and thermal processes were turned on in the GEANT4 physics lists. We use the results to simulate in detail the data obtained with different targets used in the experiment within a comprehensive analysis. This work is supported by NSF grant PHY-1307328.

  19. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system

    Science.gov (United States)

    Kahook, Samer D.; Dugan, Edward T.

    1991-01-01

    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  20. The effect of different boiling and filtering devices on the concentration of disinfection by-products in tap water.

    Science.gov (United States)

    Carrasco-Turigas, Glòria; Villanueva, Cristina M; Goñi, Fernando; Rantakokko, Panu; Nieuwenhuijsen, Mark J

    2013-01-01

    Disinfection by-products (DBPs) are ubiquitous contaminants in tap drinking water with the potential to produce adverse health effects. Filtering and boiling tap water can lead to changes in the DBP concentrations and modify the exposure through ingestion. Changes in the concentration of 4 individual trihalomethanes (THM4) (chloroform (TCM), bromodichloromethane (BDCM), dibromochloromethane (DBCM), and bromoform (TBM)), MX, and bromate were tested when boiling and filtering high bromine-containing tap water from Barcelona. For filtering, we used a pitcher-type filter and a household reverse osmosis filter; for boiling, an electric kettle, a saucepan, and a microwave were used. Samples were taken before and after each treatment to determine the change in the DBP concentration. pH, conductivity, and free/total chlorine were also measured. A large decrease of THM4 (from 48% to 97%) and MX concentrations was observed for all experiments. Bromine-containing trihalomethanes were mostly eliminated when filtering while chloroform when boiling. There was a large decrease in the concentration of bromate with reverse osmosis, but there was a little effect in the other experiments. These findings suggest that the exposure to THM4 and MX through ingestion is reduced when using these household appliances, while the decrease of bromate is device dependent. This needs to be considered in the exposure assessment of the epidemiological studies.

  1. The Effect of Different Boiling and Filtering Devices on the Concentration of Disinfection By-Products in Tap Water

    Directory of Open Access Journals (Sweden)

    Glòria Carrasco-Turigas

    2013-01-01

    Full Text Available Disinfection by-products (DBPs are ubiquitous contaminants in tap drinking water with the potential to produce adverse health effects. Filtering and boiling tap water can lead to changes in the DBP concentrations and modify the exposure through ingestion. Changes in the concentration of 4 individual trihalomethanes (THM4 (chloroform (TCM, bromodichloromethane (BDCM, dibromochloromethane (DBCM, and bromoform (TBM, MX, and bromate were tested when boiling and filtering high bromine-containing tap water from Barcelona. For filtering, we used a pitcher-type filter and a household reverse osmosis filter; for boiling, an electric kettle, a saucepan, and a microwave were used. Samples were taken before and after each treatment to determine the change in the DBP concentration. pH, conductivity, and free/total chlorine were also measured. A large decrease of THM4 (from 48% to 97% and MX concentrations was observed for all experiments. Bromine-containing trihalomethanes were mostly eliminated when filtering while chloroform when boiling. There was a large decrease in the concentration of bromate with reverse osmosis, but there was a little effect in the other experiments. These findings suggest that the exposure to THM4 and MX through ingestion is reduced when using these household appliances, while the decrease of bromate is device dependent. This needs to be considered in the exposure assessment of the epidemiological studies.

  2. Reactor-specific spent fuel discharge projections: 1986 to 2020

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs.

  3. Continuous production of Cu2ZnSnS4 nanocrystals in a flow reactor.

    Science.gov (United States)

    Shavel, Alexey; Cadavid, Doris; Ibáñez, Maria; Carrete, Alex; Cabot, Andreu

    2012-01-25

    A procedure for the continuous production of Cu(2)ZnSnS(4) (CZTS) nanoparticles with controlled composition is presented. CZTS nanoparticles were prepared through the reaction of the metals' amino complexes with elemental sulfur in a continuous-flow reactor at moderate temperatures (300-330 °C). High-resolution transmission electron microscopy and X-ray diffraction analysis showed the nanocrystals to have a crystallographic structure compatible with that of the kesterite. Chemical characterization of the materials showed the presence of the four elements in each individual nanocrystal. Composition control was achieved by adjusting the solution flow rate through the reactor and the proper choice of the nominal precursor concentration within the flowing solution. Single-particle analysis revealed a composition distribution within each sample, which was optimized at the highest synthesis temperatures used.

  4. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  5. Surface activity and radiation field measurements of the TMI-2 reactor building gross decontamination experiment

    Energy Technology Data Exchange (ETDEWEB)

    McIsaac, C V

    1983-10-01

    Surface samples were collected from concrete and metal surfaces within the Three Mile Island Unit 2 Reactor Building on December 15 and 17, 1981 and again on March 25 and 26, 1982. The Reactor Building was decontaminated by hydrolasing during the period between these dates. The collected samples were analyzed for radionuclide concentration at the Idaho National Engineering Laboratory. The sampling equipment and procedures, and the analysis methods and results are discussed. The measured mean surface concentrations of /sup 137/Cs and /sup 90/Sr on the 305-ft elevation floor before decontamination were, respectively, 3.6 +- 0.9 and 0.17 +- 0.04 ..mu..Ci/cm/sup 2/. Their mean concentrations on the 347-ft elevation floor were about the same. On both elevations, walls were found to be considerably less contaminated than floors. The fractions of the core inventories of /sup 137/Cs, /sup 90/Sr, and /sup 129/I deposited on Reactor Building surfaces prior to decontamination were calculated using their mean concentrations on various types of surfaces. The calculated values for these three nuclides are 3.5 +- 0.4 E-4, 2.4 +- 0.8 E-5, and 5.7 +- 0.5 E-4, respectively. The decontamination operations reduced the /sup 137/Cs surface activity on the 305- and 347-ft elevations by factors of 20 and 13, respectively. The /sup 90/Sr surface activity reduction was the same for both floors, that being a factor of 30. On the whole, decontamination of vertical surfaces was not achieved. Beta and gamma exposure rates that were measured during surface sampling were examined to determine the degree to which they correlated with measured surface activities. The data were fit with power functions of the form y = ax/sup b/. As might be expected, the beta exposure rates showed the best correlation. Of the data sets fit with the power function, the set of December 1981 beta exposure exhibited the least scatter. The coefficient of determination for this set was calculated to be 0.915.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Utilization of the 250 kW TRIGA Mark II reactor in Ljubljana. Thirty years of experiences

    Energy Technology Data Exchange (ETDEWEB)

    Dimic, V. [J. Stefan Institute, Ljubljana (Slovenia)

    1996-07-01

    In its 30{sup th} year, the TRIGA Mark II 250 kW pulsing reactor is continuing its busy operation. With the maximum neutron flux in the central thimble of 1.10{sup 13} n/cm{sup 2} sec and many sample radiation positions the reactor has been used for a number of sophisticated experiments in the following fields: solid state physics (elastic and inelastic scattering of neutrons), neutron dosimetry, neutron radiography, reactor physics including nuclear burn up measurements and calculations and neutron activation analysis which represents one of the major usage of our reactor. Besides these, applied research around the reactor has been conducted, such as dopping of silicon monocrystals, a routine production of various radioactive isotopes for industry and medical use ({sup 18}F,99{sup m}Tc). At the Nuclear Training Centre the TRIGA reactor is the main teaching equipment. This training centre can fulfil the training requirements of the first Slovenian Nuclear Power Plant Krsko. (orig.)

  8. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R [ORNL; Powers, Jeffrey J [ORNL; Mueller, Don [ORNL; Patton, Bruce W [ORNL

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report 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. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE.

  9. Experiences on Grid production for Geant4

    CERN Document Server

    Ribon, Alberto

    2006-01-01

    Geant4 is a general purpose toolkit for simulating the tracking and interaction of particles through matter. It is currently used in production in several particle physics experiments (BaBar, HARP, ATLAS, CMS, LHCb), and it has also applications in other areas, as space science, medical applications, and radiation studies. The complexity of the Geant4 code requires careful testing of all of its components, especially before major releases (which happens twice a year, in June and December). In this talk, I will describe the recent development of an automatic suite for testing hadronic physics in high energy calorimetry applications. The idea is to use a simplified set of hadronic calorimeters, with different beam particle types, and various beam energies, and comparing relevant observables between a given reference version of Geant4 and the new candidate one. Only those distributions that are statistically incompatible are then printed out and finally inspected by a person to look for possible bugs. The suite ...

  10. Taking a fresh Look at boiling heat transfer on the road to improved nuclear economics and efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Baglietto, E.; Pointer, W. D.

    2016-08-01

    In the effort to reinvigorate innovation in the way we design, build, and operate the nuclear power generating stations of today and tomorrow, nothing can be taken for granted. Not even the seemingly familiar physics of boiling water. The Consortium for the Advanced Simulation of Light Water Reactors, or CASL, is focused on the deployment of advanced modeling and simulation capabilities to enable the nuclear industry to reduce uncertainties in the prediction of multi-physics phenomena and continue to improve the performance of todays light water reactors and their fuel. An important part of the CASL mission is the development of a next generation thermal hydraulics simulation capability, integrating the history of engineering models based on based on experimental experience with the computing technology of the future. (Author)

  11. LMFBR type reactor and power generation system using the same

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira.

    1994-02-25

    A reactor core void reactivity of a reactor main body is set to negative or zero. A heat insulation structure is disposed on the inner wall surface of a reactor container. Oxide fuels or nitride fuels are used. A fuel pin cladding tube has a double walled structure having an outer side of stainless steel and an inner side of niobium alloy. Upon imaginary event, boiling is allowed. Even if boiling of coolants should occur by temperature elevation of fuels upon imaginary event, since reactor core fuels comprises oxides or nitrides, they have a heat resistance, further, and since the fuel pin cladding tube has super heat resistance, it has a high temperature strength, so that it is not ruptured and durable to the coolant boiling temperature. Since the reactor core void reactivity is negative or zero, the reactor core is in a subcritical state by the boiling, and the reactor core power is reduced to several % of the rated power. Accordingly, boiling and non-boiling are repeated substantially permanently in the reactor core, during which safety can be kept with no operator's handling. Further, heat generated in the reactor core is gradually removed by an air cooling system for the reactor container. (N.H.).

  12. High-speed infrared thermography for the measurement of microscopic boiling parameters on micro- and nano-structured surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Park, Youngjae; Kim, Hyungdae [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Hyungmo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Joonwon [POSTECH, Pohang (Korea, Republic of)

    2014-10-15

    Micro- and nano-scale structures on boiling surfaces can enhance nucleate boiling heat transfer coefficient (HTC) and critical heat flux (CHF). A few studies were conducted to explain the enhancements of HTC and CHF using the microscopic boiling parameters. Quantitative measurements of microscopic boiling parameters are needed to understand the physical mechanism of the boiling heat transfer augmentation on structured surfaces. However, there is no existing experimental techniques to conveniently measure the boiling parameters on the structured surfaces because of the small (boiling on micro- and nano-structured surfaces. The visualization results are analyzed to obtain the microscopic boiling parameters. Finally, quantitative microscopic boiling parameters are used to interpret the enhancement of HTC and CHF. In this study, liquid-vapor phase distributions of each surface were clearly visualized by IR thermography during the nucleate boiling phenomena. From the visualization results, following microscopic boiling parameters were quantitatively measured by image processing. - Number density of dry patch, NDP IR thermography technique was demonstrated by nucleate pool boiling experiments with M- and N surfaces. The enhancement of HTC and CHF could be explained by microscopic boiling parameters.

  13. Development of nuclear thermal hydraulic verification test and evaluation technology; study on 3-dimension measurement of two-phase flow parameters in subcooled boiling flow

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Kim, Moon Oh; Cho, Hyung Kyoo; Kim, Seong Jin [Seoul National University, Seoul (Korea)

    2002-04-01

    In this study, the experiments were conducted at different levels of inlet subcooling, flow rate and heat flux in a vertical concentric annulus channel located heater at the center with subcooled boiling conditions of atmosphere pressure and superficial velocity under 1.5m/s. The profiles of void fraction, vapor size, vapor frequency, vapor velocity and IAC were measured by 2 sensor conductivity probe in axially 3 points (L/D{sub h}=90.5,80.1,71.4) and those of liquid velocity by pitot tube. Based on the experiment data subcooled boiling models in MARS and multidimensional code, CFX-4.2 were evaluated was verified for analysis ability of these codes in subcooled boiling. 61 refs., 41 figs., 11 tabs. (Author)

  14. Rising and boiling of a drop of volatile liquid in a heavier one: application to the LMFBR severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Pigny, Sylvain L.; Coste, Pierre F. [DEN/DER/SSTH, CEA/Grenoble, 38054 Grenoble Cedex 9 (France)

    2005-07-01

    Full text of publication follows: The rising and, simultaneously the boiling, of a droplet of volatile liquid in a heavier one is computation-ally investigated. Our calculations are performed with the help of the SIMMER code, in which a specific DNS algorithm is developed, to represent surface tension between the different media in an explicit way. This is required to represent the physical contact that occurs between two liquids and the vapor from the lighter one, since interfacial heat transfers, and therefore boiling kinetics, merely depend on it. The behavior of the three fluids system is of interest as a key phenomenon related to the transition phase of LMFBR severe accidents, before the formation of a fully developed bubble column. The driven force due to the boiling of steel drops can play a major role in the relocation, and, consequently, the recriticality of UO{sub 2} fuel. The problem is investigated focusing first on analytical experiments, built-up with simulating materials, and for which accurate experimental results are provided. The dependence of results with regard to thermodynamical and physical properties is underlined. This point is of interest in view of some uncertainties in the knowledge of data concerning the materials present in the reactor at high temperature. The pressure level is a key parameter in the accident scenarios: its influence is uppermost on the volumic mass of the gas. It is also outlined. (authors)

  15. Shedding light on LMA-Dark solar neutrino solution by medium baseline reactor experiments: JUNO and RENO-50

    CERN Document Server

    Bakhti, Pouya

    2014-01-01

    In the presence of Non-Standard neutral current Interactions (NSI) a new solution to solar neutrino anomaly with $\\cos 2\\theta_{12}<0$ appears. We show that this solution can be tested by upcoming intermediate baseline reactor experiments JUNO and RENO-50.

  16. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS-8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using relevant and well-documented critical configurations from commercial pressurized water reactors. The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SCALE-4 SAS2H analytical sequence. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code family was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all calculations. This volume of the report documents a reactor critical calculation for GPU Nuclear Corporation's Three Mile Island Unit 1 (TMI-1) during hot, zero-power startup testing for the beginning of cycle 5. This unit and cycle were selected because of their relevance in spent fuel benchmark applications: (1) cycle 5 startup occurred after an especially long downtime of 6.6 years; and (2) the core consisted primarily

  17. 无筛板沸腾氯化与熔盐氯化生产TiCl4工艺浅析%The Discussion of TiCI4 Production Process of Boiling Chlorination without Sieve Plates and Melting Salt Chlorination

    Institute of Scientific and Technical Information of China (English)

    阎守义; 刘禹明

    2013-01-01

    无筛板沸腾氯化法与熔盐氯化法是目前生产TiCl4的两种主要工艺,分别是我国和前苏联钛冶金工作者针对各自地区所拥有的可用于TiCl4生产的原料而独创的生产方法,均对世界钛冶金工业的发展做出了巨大贡献.首先简要概述了两种氯化法生产TiCl4的特点,对比分析了两种方法每生产1t粗TiCl4所消耗的原料和产生的废料,以及各自的优缺点,指出采用无筛板沸腾氯化法生产的粗TiCl4的质量不如熔盐氯化法,我国的无筛板沸腾氯化法仍需继续改进.%Boiling chlorination without sieve plates and melting salt chlorination are the two main processes for TiCl4 production, which all have made great contributions to the development of titanium metallurgical industry in the world. The characteristics of the two chlorination methods were firstly summarized, then the index between boiling chlorination without sieve plates and melting salt chlorination for the production of 1 t crude TiCl4 was compared, and the strong and weak points were analyzed. Finally, it was found that the quality of crude TiCl4 produced by molten salt chlorination was better than boiling chlorination without sieve plates, and the boiling chlorination without sieve plates created by Chinese titanium metallurgy workers still needed to be improved.

  18. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The

  19. Transient measurement of temperature oscillation during noisy film boiling in superfluid helium II

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Noisy film boiling, which is characterized by a loud noise andsevere mechanical vibration, is a particular phenomenon of superfluid helium II (He II). Experiments have been conducted under various thermal conditions by varying the heating time th and the heat flux q, and the temperature oscillation during noisy film boiling is measured by the superconductor temperature sensors in order to understand the physical mechanism of noisy film boiling.

  20. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  1. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  2. Constraint on Neutrino Decay with Medium-Baseline Reactor Neutrino Oscillation Experiments

    CERN Document Server

    Abrahao, Thamys; Nunokawa, Hiroshi; Quiroga, Alexander A

    2015-01-01

    The experimental bound on lifetime of nu_3, the neutrino mass eigenstate with the smallest nu_e component, is much weaker than those of nu_1 and nu_2 by many orders of magnitude to which the astrophysical constraints apply. We argue that the future reactor neutrino oscillation experiments with medium-baseline (~ 50 km), such as JUNO or RENO-50, has the best chance of placing the most stringent constraint on nu_3 lifetime among all neutrino experiments which utilize the artificial source neutrinos. Assuming decay into invisible states, we show by a detailed chi^2 analysis that the nu_3 lifetime divided by its mass, tau_3/ m_3, can be constrained to be tau_3/m_3 > 7.5 (5.5) x 10^{-11} s/eV at 95% (99%) C.L. by 100 kt.years exposure by JUNO. It may be further improved to the level comparable to the atmospheric neutrino bound by its longer run.

  3. Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system.

    Science.gov (United States)

    Kuşçu, Özlem Selçuk; Sponza, Delia Teresa

    2011-03-15

    A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR.

  4. Experiments prior to construction of the Rapsodie reactor (1962); Experiences preliminaires a la construction de la pile rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L.; Zaleski, C.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    Before proceeding to the construction of the various reactor components described in the paper 'Fast Breeder Reactor Rapsodie', many experimental studies of a hydraulic, thermal and mechanical character have been carried out, or are under consideration, to test the validity of the principles adopted in the Preliminary Project. This paper deals with the most important of these: 1. Studies of coolant circuit components: sodium pumps (mechanical or electromagnetic), Na-NaK and NaK ir heat exchangers, measuring instruments (flow rates, temperatures), sodium purification circuits, etc. 2. Studies in cooling of fuel and fertile assemblies: a) study of the sodium cooling carried out by means of hydraulic mockups (scale of 1: 1 or over) reproducing the flow of the coolant fluid in the piping, upstream from and inside the fuel and fertile elements. b) study of the cooling by gas and by immersion in lead, employed during handling and storage operations. 3. Studies of special reactor devices: fusible rotating linkage, parts of the control rod mechanisms. 4. Study of the reactor block and coolant circuits as a whole. This study is to begin at the end of the year. The mock-up, now nearing completion, reproduces on a scale of 1: 1 the installation provided in the Preliminary Project and includes: the reactor block, to which is connected a high flow ate sodium circuit, permitting of long-term tests and thermal shocks, and also, a control rod testing circuit; complete installation of the 1 MW and 10 MW coolant circuits, the performances of which it will be possible to check under various operational conditions. 5. A safety study carried out on a 3: 10 scale mock p comprising the whole of the reactor block and shielding, with the object of limiting the effects of any accidental liberation of energy of an explosive character. (authors) [French] Avant d'entreprendre la realisation des divers elements du reacteur decrit dans le rapport 'Reacteur rapide

  5. Reactor Neutrinos

    OpenAIRE

    Lasserre, T.; Sobel, H.W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrin...

  6. Measurement of neutron energy spectrum at the radial channel No. 4 of the Dalat reactor

    OpenAIRE

    Son, Pham Ngoc; Tan, Vuong Huu

    2016-01-01

    Introduction Several compositions of neutron filters have been installed at the channel No. 4 of the Dalat research reactor to produce quasi-monoenergetic neutron beams. However, this neutron facility has been proposed to enhance the quality of the experimental instruments, and to characterize the neutron spectrum parameters for new filtered neutron beams of 2 keV, 24 keV, 59 keV and 133 keV. Case description In order to meet the demand of neutron spectrum information for calculation and desi...

  7. Experimental Study of Interfacial Friction in NaBH{sub 4} Solution in Microchannel Dehydrogenation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Seok Hyun; Hwang, Sueng Sik; Lee, Hee Joon [Kookmin Univ., Seoul (Korea, Republic of)

    2014-02-15

    Sodium borohydride (NaBH{sub 4}) is considered as a secure metal hydride for hydrogen storage and supply. In this study, the interfacial friction of two-phase flow in the dehydrogenation of aqueous NaBH{sub 4} solution in a microchannel with a hydraulic diameter of 461 μm is investigated for designing a dehydrogenation chemical reactor flow passage. Because hydrogen gas is generated by the hydrolysis of NaBH{sub 4} in the presence of a ruthenium catalyst, two different flow phases (aqueous NaBH{sub 4} solution and hydrogen gas) exist in the channel. For experimental studies, a microchannel was fabricated on a silicon wafer substrate, and 100-nm ruthenium catalyst was deposited on three sides of the channel surface. A bubbly flow pattern was observed. The experimental results indicate that the two-phase multiplier increases linearly with the void fraction, which depends on the initial concentration, reaction rate, and flow residence time.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S.; Basu, S. (Nuclear Engineering Division); (NRC)

    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

  9. Handbook of nuclear engineering: vol 1: nuclear engineering fundamentals; vol 2: reactor design; vol 3: reactor analysis; vol 4: reactors of waste disposal and safeguards

    CERN Document Server

    2013-01-01

    The Handbook of Nuclear Engineering is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all academic levels, this five volume set provides the latest findings in nuclear data and experimental techniques, reactor physics, kinetics, dynamics and control. Readers will also find a detailed description of data assimilation, model validation and calibration, sensitivity and uncertainty analysis, fuel management and cycles, nuclear reactor types and radiation shielding. A discussion of radioactive waste disposal, safeguards and non-proliferation, and fuel processing with partitioning and transmutation is also included. As nuclear technology becomes an important resource of non-polluting sustainable energy in the future, The Handbook of Nuclear Engineering is an excellent reference for practicing engineers, researchers and professionals.

  10. Effect of Mixing Driven by Siphon Flow: Parallel Experiments Using the Anaerobic Reactors with Different Mixing Modes

    Directory of Open Access Journals (Sweden)

    Kai-Qin Xu

    2013-08-01

    Full Text Available The effect of mixing by siphon flow on anaerobic digestion, sludge distribution and microbial community were examined in parallel experiments using a siphon-mixed reactor (SMR, an unmixed reactor (UMR and a continuously mixed reactor (CMR. The SMR performed well without the accumulation of fatty acids under COD loading rates varying from 3 to 18 kg/m3/day, while the UMR was totally acidified when the loading rate increased to 10 kg/m3/day. The methane yield of the SMR was at least 10% higher than that of the UMR, and comparable to that of the CMR. Furthermore, the SMR was found to markedly improve the dispersion of solids and reduce deposit formation compared to the UMR. Besides, during stable operation, the fatty acids level in the effluent of the SMR and UMR was lower than that in the CMR, and the archaeal community structure of the SMR was similar to that of the UMR.

  11. Thin film growths of zinc telluride and cadmium telluride on various substrates using a novel close space sublimation reactor CSS4

    Science.gov (United States)

    Marrufo, Damian

    Thin films of CdTe have been grown on CdS in a variety of methods for use in thin film photovoltaic systems. Limits to the efficiency of CdTe/CdS solar cells have been attributed to defects in the lattice that occur between the interface of CdS and CdTe due to a lattice mismatch. A close space sublimation (CSS) reactor known as the CSS4 was designed and fabricated in UTEP to deposit complex layers of CdTe and ZnTe on top of a CdS film that is grown via chemical bath deposition in order to obtain a CdTe photovoltaic. Unfortunately, the original design and fabrication of the CSS reactor (CSS3) proved to be unreliable and only a few depositions were made. This thesis summarizes the work done to improve the reactor to turn it into a reliable piece of lab equipment that can be used to conduct graduate level research while also listing known issues that should be addressed in future modifications. A standard procedure for growths is also presented. Some early results of films and devices made with the new CSS4 reactor are also included to demonstrate the potential experiments that could be conducted and the type of results that can be expected.

  12. Experiments on rehabilitation of radioactive metallic waste (RMW) of reactor stainless steels of Siberian chemical plant

    Science.gov (United States)

    Kolpakov, G. N.; Zakusilov, V. V.; Demyanenko, N. V.; Mishin, A. S.

    2016-06-01

    Stainless steel pipes, used to cool a reactor plant, have a high cost, and after taking a reactor out of service they must be buried together with other radioactive waste. Therefore, the relevant problem is the rinse of pipes from contamination, followed by returning to operation.

  13. Air purification in a reverse-flow reactor: Model simulations vs. experiments

    NARCIS (Netherlands)

    Beld, van de L.; Westerterp, K.R.

    1996-01-01

    The behavior of a reverse-flow reactor was studied for the purification of polluted air by catalytic combustion. A heterogeneous one-dimensional model was extended with a heat balance for the reactor wall. An overall heat transport term is included to account for the small heat losses in radial dire

  14. Novel Gas-liquid Hybrid Discharge Reactor for 4-CP Containing Wastewater Treatment

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In order to improve 4-CP degradation efficiency, a novel gas-liquid hybrid discharge (HD) reactor was developed. Removal of 4-CP with spark-spark discharge (SSD) was higher than that with spark-corona discharge (SCD). Amount of H2O2 and O3 produced with SSD were larger than that with SCD. ·OH formation was increased by the combination of H2O2 and O3.The contribution of ·OH (38 % formed by O3 conversion) oxidation on removal of 4-CP accounted for nearly 60 %. The other effects of ultraviolet radiation, intense shock waves and pyrolysis,played partial roles in about 40 % of removal rate.

  15. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  16. Study of rapid transient explosive boiling under short-pulsed laser heating

    Institute of Scientific and Technical Information of China (English)

    JIN Renxi; HUAI Xiulan; LIU Dengying

    2004-01-01

    Using acetone, ethanol, water and acetone-water mixture as test liquids, the rapid transient explosive boiling (RTEB) taking place under short-pulsed laser heating is observed in experiment. The behaviors of temperature variation are investigated via transient temperature measurement. The vapor bubble behaviors of RTEB are captured by high-speed photography, and the difference between RTEB and normal boiling is analyzed. The boiling heat transfer of RTEB is also discussed. It shows that the character of RTEB is far different from normal boiling.

  17. Experimental Evidence of the Vapor Recoil Mechanism in the Boiling Crisis

    CERN Document Server

    Nikolayev, Vadim; Garrabos, Y; Beysens, D

    2016-01-01

    Boiling crisis experiments are carried out in the vicinity of the liquid-gas critical point of H2. A magnetic gravity compensation setup is used to enable nucleate boiling at near critical pressure. The measurements of the critical heat flux that defines the threshold for the boiling crisis are carried out as a function of the distance from the critical point. The obtained power law behavior and the boiling crisis dynamics agree with the predictions of the vapor recoil mechanism and disagree with the classical vapor column mechanism.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, F

    2003-06-01

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

  20. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    Science.gov (United States)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  1. Surface boiling of superheated liquid

    Energy Technology Data Exchange (ETDEWEB)

    Reinke, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-01-01

    A basic vaporization mechanism that possibly affects the qualitative and quantitative prediction of the consequences of accidental releases of hazardous superheated liquids was experimentally and analytically investigated. The studies are of relevance for the instantaneous failure of a containment vessel filled with liquefied gas. Even though catastrophical vessel failure is a rare event, it is considered to be a major technological hazard. Modeling the initial phase of depressurisation and vaporization of the contents is an essential step for the subsequent analysis of the spread and dispersion of the materials liberated. There is only limited understanding of this inertial expansion stage of the superheated liquid, before gravity and atmospheric turbulence begin to dominate the expansion. This work aims at a better understanding of the vaporization process and to supply more precise source-term data. It is also intended to provide knowledge for the prediction of the behavior of large-scale releases by the investigation of boiling on a small scale. Release experiments with butane, propane, R-134a and water were conducted. The vaporization of liquids that became superheated by sudden depressurisation was studied in nucleation-site-free glass receptacles. Several novel techniques for preventing undesired nucleation and for opening the test-section were developed. Releases from pipes and from a cylindrical geometry allowed both linear one-dimensional, and radial-front two-dimensional propagation to be investigated. Releases were made to atmospheric pressure over a range of superheats. It was found that, above a certain superheat temperature, the free surface of the metastable liquid rapidly broke up and ejected a high-velocity vapor/liquid stream. The zone of intense vaporization and liquid fragmentation proceeded as a front that advanced into the test fluids. No nucleation of bubbles in the bulk of the superheated liquid was observed. (author) figs., tabs., refs.

  2. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    Energy Technology Data Exchange (ETDEWEB)

    Shah, Vikram Naginbhai; Ware, Arthur Gates; Atwood, Corwin Lee; Sattison, Martin Blaine; Hartley, Robert Scott; Hsu, C.

    1999-08-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number of rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

  3. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    Energy Technology Data Exchange (ETDEWEB)

    A. G. Ware; C. Hsu (USNRC); C. L. Atwood; M. B. Sattison; R. S. Hartley (INEEL); V. N. Shah

    1999-02-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number and rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

  4. A descriptive model of the molten salt reactor experiment after shutdown: Review of FY 1995 progress

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.; Del Cul, G.D.; Toth, L.M.

    1996-01-01

    During FY 1995 considerable progress was made toward gaining a better understanding of the chemistry and transport processes that continue to govern the behavior of the Molten Salt Reactor Experiment (MSRE). As measurements in the MSRE proceed, laboratory studies continue, and better analyses are available, our understanding of the state of the MSRE and the best path toward remediation improves. Because of the immediate concern about the deposit in the auxiliary charcoal bed (ACB), laboratory studies in the past year focused on carbon-fluorine chemistry. Secondary efforts were directed toward investigation of gas generation from MSRE salts by both radiolytic and nonradiolytic pathways. In addition to the laboratory studies, field measurements at the MSRE provided the basis for estimating the inventory of uranium and fluorine in the ACB. Analysis of both temperature and radiation measurements provided independent and consistent estimates of about 2.6 kg of uranium deposited in the top of the ACB. Further analysis efforts included a refinement in the estimates of the fuel- salt source term, the deposited decay energy, and the projected rate of radiolytic gas generation. This report also provides the background material necessary to explain new developments and to review areas of particular interest. The detailed history of the MSRE is extensively documented and is cited where appropriate. This work is also intended to update and complement the more recent MSRE assessment reports.

  5. Thermal analysis to support decommissioning of the molten salt reactor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Sulfredge, C.D.; Morris, D.G.; Park, J.E.; Williams, P.T.

    1996-06-01

    As part of the decommissioning process for the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory, several thermal-sciences issues were addressed. Apparently a mixture of UF{sub 6} and F{sub 2} had diffused into the upper portion of one charcoal column in the MSRE auxiliary charcoal bed (ACB), leading to radiative decay heating and possible chemical reaction sources. A proposed interim corrective action was planned to remove the water from the ACB cell to reduce criticality and reactivity concerns and then fill the ACB cell with an inert material. This report describes design of a thermocouple probe to obtain temperature measurements for mapping the uranium deposit, as well as development of steady-state and transient numerical models for the heat transfer inside the charcoal column. Additional numerical modeling was done to support filling of the ACB cell. Results from this work were used to develop procedures for meeting the goals of the MSRE Remediation Project without exceeding appropriate thermal limits.

  6. Isothermal flow measurement using planar PIV in the 1/4 scaled model of CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Im, Sunghyuk; Sung, Hyung Jin [KAIST, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the flow field and local temperature distribution in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) techniques. Small-scale 1/40 and 1/8 small-scale model tests were performed prior to installation of the main MCT facility to identify the potential problems of the flow visualization and measurement expected in the 1/4 scale MCT facility. In the 1/40 scale test, a flow field was measured with a PIV measurement technique under an iso-thermal state, and the temperature field was visualized using a LIF technique. In this experiment, the key point was to illuminate the region of interest as uniformly as possible since the velocity and temperature fields in the shadow regions were distorted and unphysical. In the 1/8 scale test, the flow patterns from the inlet nozzles to the top region of the tank were investigated using PIV measurement at two different positions of the inlet nozzle. For each position of laser beam exposure the measurement sections were divided to 7 groups to overcome the limitation of the laser power to cover the relatively large test section. The MCT facility is the large-scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. It is reduced in a 1/4 scale and a moderator test vessel is built to the specifications of the CANDU6 reactor design, where a working fluid is sub-cooled water with atmospheric pressure. Previous studies were

  7. Subtask 12G3: Fracture properties of V-4Cr-4Ti irradiated in the dynamic helium charging experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Nowicki, L.J.; Busch, D.E.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    The objective of this work is to determine the effect of simultaneous displacement damage and dynamically charged helium on the ductile-brittle transition behavior of V-4Cr-4Ti specimens irradiated to 18-31 dpa at 425-600{degrees}C in the Dynamic Helium Charging Experiment (DHCE). One property of vanadium-base alloys that is not well understood in terms of their potential use as fusion reactor structural materials is the effect of simultaneous generation of helium and neutron damage under conditions relevant to fusion reactor operation. In the present DHCE, helium was produced uniformly in the specimen at linear rates ranging from {approx}0.4 to 4.2 appm helium/dpa by the decay of tritium during irradiation to 18-31 dpa at 425-600{degrees}C in Li-filled DHCE capsules in the Fast Flux Test Facility. Ductile-brittle transition behavior of V-4Cr-4Ti, recently identified as the most promising vanadium-base alloy for fusion reactor use, was determined from multiple-bending tests (at -196{degrees}C to 50{degrees}C) and quantitative SEM fractography on TEM disks (0.3-mm thick) and broken tensile specimens (1.0-mm thick). No brittle behavior was observed at temperatures >-150{degrees}C, and predominantly brittle-cleavage fracture morphologies were observed only at -196{degrees}C in some specimens irradiated to 31 dpa at 425{degrees}C during DHCE. Ductile-brittle transition temperatures (DBTTs) were -200{degrees}C to -175{degrees}C for both types of specimens. In strong contrast to tritium-trick experiments in which dense coalescence of helium bubbles is produced on grain boundaries in the absence of displacement damage, no intergranular fracture was observed in the bend-tested specimens irradiated in the DHCE. 24 refs., 3 figs., 2 tabs.

  8. Resting Study of Tracer Experiment on Catalytic Wet Oxidation Reactor under Micro-gravity and Earth Gravity Conditions

    Institute of Scientific and Technical Information of China (English)

    YANG Ji; JIA Jin-ping

    2005-01-01

    The International Space Station(ISS) employs catalytic wet oxidation carried out in a Volatile Reactor Assembly (VRA) for water recycling. Previous earth gravity experiments show that the VRA is very effective at removing polar,low molecular weight organics. To compare the reactor performance under micro-gravity and Earth gravity conditions,a tracer study was performed on a space shuttle in 1999 by using 0. 2% potassium carbonate as the chemical tracer.In this paper, the experimental data were analyzed and it is indicated that the reactor can be considered as a plug flow one under both micro-gravity and earth gravity experimental conditions. It has also been proved that dispersion is not important in the VRA reactor under the experimental conditions. Tracer retardation was observed in the experiments and it is most likely caused by catalyst adsorption. It is concluded that the following reasons may also have influence on the retardation of mean residence time: (1) the liquid can be held by appurtenances, which will retard the mean residence time; (2) the pores can hold the tracer, which can also retard the mean residence time.

  9. Remediation of Water Contaminated with an Azo Dye: An Undergraduate Laboratory Experiment Utilizing an Inexpensive Photocatalytic Reactor

    Science.gov (United States)

    Bumpus, John A.; Tricker, Jennifer; Andrzejewski, Ken; Rhoads, Heather; Tatarko, Matthew

    1999-12-01

    The construction and use of an inexpensive photocatalytic reactor that utilizes titanium dioxide as the photocatalyst for wastewater treatment is described. In these experiments and in supplementary material, students are made aware that a variety of techniques have been developed to treat wastewaters, including those generated by the chemical industry. Water contaminated with the azo dye Congo Red was selected as an example of how one might treat contaminated water from a textile manufacturing facility. These experiments emphasize that, in addition to product development, chemists must also be concerned with waste treatment. A summary of the theory of titanium dioxide-mediated photocatalysis is provided. The phenomenon of photosensitization is also discussed. The usefulness of Congo Red is summarized and a brief history of this dye is given. In addition to being inexpensive, the photocatalytic reactor described is easy to construct and uses a readily available low-wattage fluorescent light. An important feature of this reactor is that the heat generated by the light source is readily dissipated by the water undergoing treatment. Thus no special cooling apparatus is required. One of the most important aspects of this work is that it provides a wide variety of continuing research suggestions that would be suitable and readily accomplished in undergraduate departments and high school laboratories; even those where budgetary priorities are a major concern. Use of this reactor would also enable students to design systems to treat "real-world" wastes, including some that are generated in instructional laboratories.

  10. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  11. Student Collaboration in a Series of Integrated Experiments to Study Enzyme Reactor Modeling with Immobilized Cell-Based Invertase

    Science.gov (United States)

    Taipa, M. A^ngela; Azevedo, Ana M.; Grilo, Anto´nio L.; Couto, Pedro T.; Ferreira, Filipe A. G.; Fortuna, Ana R. M.; Pinto, Ine^s F.; Santos, Rafael M.; Santos, Susana B.

    2015-01-01

    An integrative laboratory study addressing fundamentals of enzyme catalysis and their application to reactors operation and modeling is presented. Invertase, a ß-fructofuranosidase that catalyses the hydrolysis of sucrose, is used as the model enzyme at optimal conditions (pH 4.5 and 45 °C). The experimental work involves 3 h of laboratory time…

  12. A pilot plant scale reactor/separator for ethanol from cellulosics. ERIP/DOE quarterly report no. 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Dale, M.C.; Moelhman, M.; Butters, R.

    1998-12-01

    The objective of this project is to develop and demonstrate a continuous, low energy process for the conversion of cellulosics to ethanol. This process involves a pretreatment step followed by enzymatic release of sugars and the consecutive simultaneous saccharification/fermentation (SSF) of cellulose (glucans) followed by hemi-cellulose (pentosans) in a multi-stage continuous stirred reactor separator (CSRS). During quarters 3 and 4, we have completed a literature survey on cellulase production, activated one strain of Trichoderma reesei. We continued developing our proprietary Steep Delignification (SD) process for biomass pretreatment. Some problems with fermentations were traces to bad cellulase enzyme. Using commercial cellulase enzymes from Solvay & Genecor, SSF experiments with wheat straw showed 41 g/L ethanol and free xylose of 20 g/L after completion of the fermentation. From corn stover, we noted 36 g/L ethanol production from the cellulose fraction of the biomass, and 4 g/L free xylose at the completion of the SSF. We also began some work with paper mill sludge as a cellulose source, and in some preliminary experiments obtained 23 g/L ethanol during SSF of the sludge. During year 2, a 130 L process scale unit will be operated to demonstrate the process using straw or cornstalks. Co-sponsors of this project include the Indiana Biomass Grants Program, Bio-Process Innovation.

  13. A study on the correlations development for film boiling heat transfer on spheres

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corlium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced. 7 refs., 5 figs., 5 tabs. (Author)

  14. Numerical analysis on pool boiling using user defined function

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Sung Uk; Jeon, Byong Guk; Kim, Seok; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    PAFS (passive auxiliary feedwater system) adopted in the APR+ (Advanced Power Reactor Plus) of Korea is one such application. When PAFS is activated with an actuation signal, steam from the steam generator passes through heat exchanger tubes submerged in a water tank of the PAFS. Outside these heat exchanger tubes, nucleate boiling phenomena appears. In the present work, a numerical study is reported on three-dimensional transient state pool boiling of water having an immersed heat source. The velocity vector fields during the decrease in the water level are numerically investigated in a pool, and the accuracy of the results is checked by comparing the experimental results conducted using the PIV techniques by Kim et al. These numerical results can be used as basic research data for an analysis and prediction of the natural circulation phenomena in the cooling tank of the passive safety system in a nuclear power plant.

  15. Study of film boiling collapse behavior during vapor explosion

    Energy Technology Data Exchange (ETDEWEB)

    Yagi, Masahiro; Yamano, Norihiro; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Abe, Yutaka; Adachi, Hiromichi; Kobayashi, Tomoyoshi

    1996-06-01

    Possible large scale vapor explosions are safety concern in nuclear power plants during severe accident. In order to identify the occurrence of the vapor explosion and to estimate the magnitude of the induced pressure pulse, it is necessary to investigate the triggering condition for the vapor explosion. As a first step of this study, scooping analysis was conducted with a simulation code based on thermal detonation model. It was found that the pressure at the collapse of film boiling much affects the trigger condition of vapor explosion. Based on this analytical results, basic experiments were conducted to clarify the collapse conditions of film boiling on a high temperature solid ball surface. Film boiling condition was established by flooding water onto a high temperature stainless steel ball heated by a high frequency induction heater. After the film boiling was established, the pressure pulse generated by a shock tube was applied to collapse the steam film on the ball surface. As the experimental boundary conditions, materials and size of the balls, magnitude of pressure pulse and initial temperature of the carbon and stainless steel balls were varied. The transients of pressure and surface temperature were measured. It was found that the surface temperature on the balls sharply decreased when the pressure wave passed through the film on balls. Based on the surface temperature behavior, the film boiling collapse pattern was found to be categorized into several types. Especially, the pattern for stainless steel ball was categorized into three types; no collapse, collapse and reestablishment after collapse. It was thus clarified that the film boiling collapse behavior was identified by initial conditions and that the pressure required to collapse film boiling strongly depended on the initial surface temperature. The present results will provide a useful information for the analysis of vapor explosions based on the thermal detonation model. (J.P.N.)

  16. Implementation of the interfacial area transport equation in trace for boiling two-phase flows

    Science.gov (United States)

    Bernard, Matthew S.

    Correctly predicting the interfacial area concentration (a i) is vital to the overall accuracy of the two-fluid model because ai describes the amount of surface area that exists between the two-phases, and is therefore directly related to interfacial mass, momentum and energy transfer. The conventional method for specifying ai in the two-fluid model is through flow regime-based empirical correlations coupled with regime transition criteria. However, a more physically consistent approach to predicting ai is through the interfacial area transport equation (IATE), which can address the deficiencies of the flow regime-based approach. Some previous studies have been performed to demonstrate the feasibility of IATE in developmental versions of the nuclear reactor systems analysis code, TRACE. However, a full TRACE version capable of predicting boiling two-phase flows with the IATE has not been established. Therefore, the current work develops a version of TRACE that is capable of predicting boiling two-phase flows using the IATE. The development is carried out in stages. First, a version of TRACE which employs the two-group IATE for adiabatic, vertical upward, air-water conditions is developed. An in-depth assessment on the existing experimental database is performed to select reliable experimental data for code assessment. Then, the implementation is assessed against the qualified air-water two-phase flow experimental data. Good agreement is observed between the experimental data for ai and the TRACE code with an average error of +/-9% for all conditions. Following the initial development, one-group IATE models for vertical downward and horizontal two-phase flows are implemented and assessed against qualified data. Finally, IATE models capable of predicting subcooled boiling two-phase flows are implemented. An assessment of the models shows that TRACE is capable of generating ai in subcooled boiling two-phase flows with the IATE and that heat transfer effects dominate

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh Thi-Cam [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, “Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program”). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, “AGR 3/4 Daily As Run Thermal Analyses”. The ABAQUS code’s finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, “JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle Experiment in ATR

  20. Investigation of neutronic performance of a peaceful nuclear explosive reactor (PACER) evaluating UF4 and ThF4 nuclides

    Indian Academy of Sciences (India)

    Selahaddin Orhan Akansu; Nafiz Kahraman

    2004-08-01

    In this study, neutronic behaviour of a peaceful nuclear explosion reactor (PACER) producing approx 1.2 GWe power from fusion explosions is analysed considering ThF4 and UF4 as fissile zones. UF4 and ThF4 are put in to the system adjacent to the inner-side of the flibe coolant zone positions (distance between explosive region and fuel zone, DR = 50, 100, 200, 400, 500 and 700 cm). Flibe percentages are taken to be 25, 50, 75 and 100%. It is found that optimum combinations of fissile zone thickness, coolant zone position and coolant percentages are 10 cm, 200 cm and 25% flibe for UF4 and 10 cm, 50 cm and 25% flibe for ThF4. The behaviour of fuels mentioned above has been observed over a period of 30 years at discrete time intervals, $\\Delta t = 30$ days. In the case of UF4; values at DR = 200 cm reach 5.35, 5.22, 4.88, and 4.88 from 3.12, 2.98, 2.83 and 2.83, for 25, 50, 75, 100% flibe respectively. For ThF4 at DR = 50 cm, values reach from 1.61, 1.54, 1.50 and 1.46 to 1.93, 2.00, 2.04 and 1.99 for 25, 50, 75, 100% flibe respectively. Cumulative fissile fuel breeding (CFFE) values reach up to 5.5 from 0.7 at DR = 200 and 25% case for UF4, and up to 6.36% from 0%, at DR = 50 and 25% flibe, for ThF4, at the end of the operation period.

  1. Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

    OpenAIRE

    Aldawahra Saadou; Khattab Kassem; Saba Gorge

    2015-01-01

    Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The propos...

  2. Buoyancy induced limits for nanoparticle synthesis experiments in horizontal premixed low-pressure flat-flame reactors

    Science.gov (United States)

    Weise, C.; Faccinetto, A.; Kluge, S.; Kasper, T.; Wiggers, H.; Schulz, C.; Wlokas, I.; Kempf, A.

    2013-06-01

    Premixed low-pressure flat-flame reactors can be used to investigate the synthesis of nanoparticles. The present work examines the flow field inside such a reactor during the formation of carbon (soot) and iron oxide (from Fe(CO)5) nanoparticles, and how it affects the measurements of nanoparticle size distribution. The symmetry of the flow and the impact of buoyancy were analysed by three-dimensional simulations and the nanoparticle size distribution was obtained by particle mass spectrometry (PMS) via molecular beam sampling at different distances from the burner. The PMS measurements showed a striking, sudden increase in particle size at a critical distance from the burner, which could be explained by the flow field predicted in the simulations. The simulation results illustrate different fluid mechanical phenomena which have caused this sudden rise in the measured particle growth. Up to the critical distance, buoyancy does not affect the flow, and an (almost) linear growth is observed in the PMS experiments. Downstream of this critical distance, buoyancy deflects the hot gas stream and leads to an asymmetric flow field with strong recirculation. These recirculation zones increase the particle residence time, inducing very large particle sizes as measured by PMS. This deviation from the assumed symmetric, one-dimensional flow field prevents the correct interpretation of the PMS results. To overcome this problem, modifications to the reactor were investigated; their suitability to reduce the flow asymmetry was analysed. Furthermore, 'safe' operating conditions were identified for which accurate measurements are feasible in premixed low-pressure flat-flame reactors that are transferrable to other experiments in this type of reactor. The present work supports experimentalists to find the best setup and operating conditions for their purpose.

  3. Fuel loading experiments of IPEN/MB-01 reactor core; Experimento de carregamento do nucleo do reator IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo; Angioletto, Elcio; Pasqualetto, Hertz; Moreira, Joao M.L.; Sabo, Marcos A.; Fuga, Rinaldo [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: ayabe@net.ipen.br; Jerez, Rogerio [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2000-07-01

    IPEN/MB-01 is a critical assembly which reaches a first critically at November 1998, since then a several experiments have been conducted in order to validate and qualify the reactor Physics methodologies and codes. Recently, loading experiment was performed to determine the critical number of fuels rods using a multiplication inverse technique (1/M). The experiment can be considered as benchmark and contribute toward reactor physics methodologies, codes and basic nuclear data libraries validation qualification. (author)

  4. Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

    Energy Technology Data Exchange (ETDEWEB)

    Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)

    2013-09-15

    Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)

  5. Nuclear data uncertainty quantification and data assimilation for a lead-cooled fast reactor : Using integral experiments for improved accuracy

    OpenAIRE

    Alhassan, Erwin

    2015-01-01

    For the successful deployment of advanced nuclear systems and optimization of current reactor designs, high quality nuclear data are required. Before nuclear data can be used in applications they must first be evaluated, tested and validated against a set of integral experiments, and then converted into formats usable for applications. The evaluation process in the past was usually done by using differential experimental data which was then complemented with nuclear model calculations. This t...

  6. Physical Design of Critical Experiment Facility for Verifying Characteristics and Effects of Coupling Between Reactor and Spallation Target of ADS

    Institute of Scientific and Technical Information of China (English)

    YIN; Sheng-gui; ZHOU; Qi; LI; Yan

    2013-01-01

    For the purpose of studying and verifying characteristics and effects of coupling between reactor and spallation target of ADS,based on the critical experimental facility design criteria and the availableexperiment condition,physical design of a critical experiment facility with lead coolant is completed,using critical calculation code MONK-9A.The contents of physical designs mainly include nuclear fuel,array of fuel rods,neutron source

  7. Selection of nuclear reactors through the hierarchic analysis process: the Mexican case; Seleccion de reactores nucleares mediante el proceso de analisis jerarquico: el caso Mexicano

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo, C.; Nelson, P.F.; Francois, J.L. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, 62550 Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2008-07-01

    In this work the decision making method known as hierarchical analysis process for the selection of a new reactor in Mexico was applied. The main objective of the process it is to select the nuclear reactor technology more appropriate for Mexico, to begin the bid process inside one or two years to begin their operation in 2016. The options were restricted to four reactors that fulfill the following ones approaches: 1) its are advanced reactors, from the technological point of view, with regard to the reactors that at the moment operate in the Laguna Verde Power Station, 2) its are reactors that have the totally finished design, 3) its are reactors that already have the certification on the part of the regulator organism of the origin country or that they are in an advanced state of the certification process and 4) its are reactors offered by the companies that they have designed and built the greater number of reactors that are at the moment in operation at world level. Taking into account these restrictions it was decided to consider as alternative at the reactors: Advanced Boiling Water Reactor (A BWR), European Reactor of Pressurized Water (EPR), Water at Pressure reactor (AP1000) and Simplified Economic Reactor of Boiling Water (ESBWR). The evaluation approaches include economic and of safety indicators, qualitative some of them and other quantitative ones. Another grade of complexity in the solution of the problem is that there are actors that can be involved in the definition of the evaluation approaches and in the definition of the relative importance among them, according to each actor's interests. To simplify the problem its were only considered two actors or groups of interest that can influence in more significant way and that are the Federal Commission of Electricity and the National Commission of Nuclear Safety and Safeguards. The qualifications for each reactor in function of the evaluation approaches were obtained, being the A BWR the best

  8. An Integrated Management System (IMS) for JM-1 SLOWPOKE-2 research reactor in Jamaica: experiences in documentation

    Energy Technology Data Exchange (ETDEWEB)

    Warner, T., E-mail: traceyann.warner02@uwimona.edu.jm [Univ. of West Indies, Mona (Jamaica)

    2014-07-01

    Since the first criticality in March 1984, the Jamaica SLOWPOKE-2 research reactor at the University of the West Indies, Mona located in the department of the International Centre for Environmental and Nuclear Sciences (ICENS) has operated for approximately 52% of the lifetime of the existing core configuration. The 20kW pool type research reactor has been primarily used for neutron activation analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration in Jamaica. The involvement of the JM-1 reactor for research and teaching activities has segued into commercial applications which, coupled with the current core conversion programme from HEU to LEU, has demanded the implementation of management systems to satisfy regulatory requirements and assure compliance with internationally defined quality standards. At ICENS, documentation related to the Quality Management System aspect of an Integrated Management System (IMS) is well underway. The quality system will incorporate operational and nuclear safety, training, maintenance, design, utilization, occupational health and safety, quality service, and environmental management for its Nuclear Analytical Laboratory, NAL. The IMS is being designed to meet the requirements of the IAEA GS-R-3 with additional controls from international standards including: ISO/IEC 17025:2005, ISO 9001:2008, ISO 14001:2004 and OHSAS 18001:2007. This paper reports on the experiences of the documentation process in a low power reactor facility characterized by limited human resource, where innovative mechanisms of system automation and modeling are included to increase productivity and efficiency. (author)

  9. 76 FR 31381 - Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

    Science.gov (United States)

    2011-05-31

    ... Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 4 to Standard Review Plan (SRP... Position (BTP) 8-8 on ``Onsite (Emergency Diesel Generators) and Offsite Power Sources Allowed Outage Time... COMMISSION Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

  10. Marangoni heat transfer in subcooled nucleate pool boiling

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, S.; Robinson, T.; Judd, R.L. [McMaster University, Hamilton, ON (Canada). Dept. of Mechanical Engineering

    2004-11-01

    The liquid motion induced by surface tension variation, termed the Marangoni effect, and its contribution to boiling heat transfer has been an issue of much controversy. Boiling heat transfer theory, although acknowledging its existence, considers its contribution to heat transfer to be insignificant in comparison with buoyancy induced convection. However, recent microgravity experiments have shown that although the boiling mechanism in a reduced gravity environment is different, the corresponding heat transfer rates are similar to those obtained under normal gravity conditions, raising questions about the validity of the assumption. An experimental investigation was performed in which distilled water was gradually heated to boiling conditions on a copper heater surface at four different levels of subcooling. Photographic investigation of the bubbles appearing on the surface was carried out in support of the measurements. The results obtained indicate that Marangoni convection associated with the bubbles formed by the air dissolved in the water which emerged from solution when the water was heated sufficiently, significantly influenced the heat transfer rate in subcooled nucleate pool boiling. A heat transfer model was developed in order to explain the phenomena observed. (author)

  11. Hysteresis of boiling for different tunnel-pore surfaces

    Directory of Open Access Journals (Sweden)

    Pastuszko Robert

    2015-01-01

    Full Text Available Analysis of boiling hysteresis on structured surfaces covered with perforated foil is proposed. Hysteresis is an adverse phenomenon, preventing high heat flux systems from thermal stabilization, characterized by a boiling curve variation at an increase and decrease of heat flux density. Experimental data were discussed for three kinds of enhanced surfaces: tunnel structures (TS, narrow tunnel structures (NTS and mini-fins covered with the copper wire net (NTS-L. The experiments were carried out with water, R-123 and FC-72 at atmospheric pressure. A detailed analysis of the measurement results identified several cases of type I, II and III for TS, NTS and NTS-L surfaces.

  12. High level disinfection of a home care device; to boil or not to boil?

    Science.gov (United States)

    Winthrop, K L; Homestead, N

    2012-03-01

    We developed a percutaneous electrical transducer for home therapy of chronic pain, a device that requires high level disinfection between uses. The utility of boiling water to provide high level disinfection was evaluated by inoculating transducer pads with potential skin pathogens (Staphylococcus aureus, Mycobacterium terrae, Pseudomonas aeruginosa, Candida albicans) and subjecting them to full immersion in water boiling at 4200 feet elevation (95 °C). Log10 reductions in colony-forming units (cfu) at 10 min were 7.1, >6.3 and >5.5 for S. aureus, P. aeruginosa and C. albicans, respectively, but only 4.6 for M. terrae. At 15 min the reductions had increased to 7.5, >6.8, >6.6 and >7.5 cfu, respectively.

  13. Engineering Evaluation of Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiement for the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlberg, Jon A.; Roberts, Kenneth T.; Kollie, Thomas G.; Little, Leslie E.; Brady, Sherman D.

    2009-09-30

    This evaluation was performed by Pro2Serve in accordance with the Technical Specification for an Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory (BJC 2009b). The evaluators reviewed the Engineering Evaluation Work Plan for Molten Salt Reactor Experiment Residual Salt Removal, Oak Ridge National Laboratory, Oak Ridge, Tennessee (DOE 2008). The Work Plan (DOE 2008) involves installing a salt transfer probe and new drain line into the Fuel Drain Tanks and Fuel Flush Tank and connecting them to the new salt transfer line at the drain tank cell shield. The probe is to be inserted through the tank ball valve and the molten salt to the bottom of the tank. The tank would then be pressurized through the Reactive Gas Removal System to force the salt into the salt canisters. The Evaluation Team reviewed the work plan, interviewed site personnel, reviewed numerous documents on the Molten Salt Reactor (Sects. 7 and 8), and inspected the probes planned to be used for the transfer. Based on several concerns identified during this review, the team recommends not proceeding with the salt transfer via the proposed alternate salt transfer method. The major concerns identified during this evaluation are: (1) Structural integrity of the tanks - The main concern is with the corrosion that occurred during the fluorination phase of the uranium removal process. This may also apply to the salt transfer line for the Fuel Flush Tank. Corrosion Associated with Fluorination in the Oak Ridge National Laboratory Fluoride Volatility Process (Litman 1961) shows that this problem is significant. (2) Continued generation of Fluorine - Although the generation of Fluorine will be at a lower rate than experienced before the uranium removal, it will continue to be generated. This needs to be taken into consideration regardless of what actions are taken with the salt. (3) More than one phase of material

  14. Status of the spent fuel in the reactor buildings of Fukushima Daiichi 1–4

    Energy Technology Data Exchange (ETDEWEB)

    Jäckel, Bernd S., E-mail: bernd.jaeckel@psi.ch

    2015-03-15

    The ratios of the radionuclides Cs-134g and Cs-137 deduced from measurements of liquid samples from the spent fuel pools in Fukushima Daiichi 1–4 are used to interpret the status of the spent fuel assemblies in the pools of the damaged reactor buildings. The different natures of the production of Cs-134g (neutron capture product of Cs-133) and Cs-137 (cumulative fission product from mass chain 137) and the different half-lives (2.06 years and 30.17 years respectively) require a complicated calculation of the mass and activity of the two nuclides. These masses are depending on the local burn up of the fuel, the burn up history and the radioactive decay. Calculation of the neutron capture product Cs-134g is particularly complicated, because the production of Cs-133 (stable cumulative fission product from mass chain 133) has to be taken into account. The neutron capture cross section for Cs-133 for thermal neutrons is well known, but the energy spectrum of the neutrons in a reactor includes higher energies according to the degree of moderation. Therefore the cross section was fitted from a gamma scan of spent fuel rods in a hot cell. The method of the calculation of the nuclide activities and the interpretation of the gamma measurements of the spent fuel pool samples from Fukushima Daiichi 1–4 are described in detail. It could be shown that at most only very minor mechanical damage of some spent fuel elements occurred during the accident and the later phase of the clearing work.

  15. Visualization of pool boiling from complex surfaces with internal tunnels

    Directory of Open Access Journals (Sweden)

    Pastuszko Robert

    2012-04-01

    Full Text Available The paper presents experimental investigations of boiling heat transfer for a system of connected narrow horizontal and vertical tunnels. These extended surfaces, named narrow tunnel structure (NTS, can be applied to electronic element cooling. The experiments were carried out with ethanol at atmospheric pressure. The tunnel external covers were manufactured out of 0.1 mm thick perforated copper foil (hole diameters 0.5 mm, sintered with the mini-fins, formed on the vertical side of the 10 mm high rectangular fins and horizontal inter-fin surface. Visualization studies were conducted with a transparent structured model of joined narrow tunnels limited with the perforated foil. The visualization investigations aimed to formulate assumptions for the boiling model through distinguishing boiling types and defining all phases of bubble growth.

  16. Micro-column enhanced boiling structure and its ramification

    Institute of Scientific and Technical Information of China (English)

    汤勇; 陆龙生; 袁冬; 苏达士

    2008-01-01

    Enhanced boiling experiments of two different enhanced structures were carried out in a thermosyphon loop evaporator chamber. One was micro-columns array structure (MCAS), which was fabricated on copper plate surface with interaction high speed wire electrode discharge machining (HS-WEDM). The other was the ramification of MCAS, named micro-column-array and sintered-copper compound structure (MSCS), which was fabricated with sintered method on micro-column array structure. Considering the wall superheat and critical heat flux (CHF), comparisons were made between them. The results show that both MCAS and MSCS can enhance the boiling heat transfer. It is also found that the enhanced boiling heat transfer ability of MSCS is changed obviously while the porosity of the sintered copper layer is changed.

  17. On mechanism of explosive boiling in nanosecond regime

    Science.gov (United States)

    Çelen, Serap

    2016-06-01

    Today laser-based machining is used to manufacture vital parts for biomedical, aviation and aerospace industries. The aim of the paper is to report theoretical, numerical and experimental investigations of explosive boiling under nanosecond pulsed ytterbium fiber laser irradiation. Experiments were performed in an effective peak power density range between 1397 and 1450 MW/cm2 on pure titanium specimens. The threshold laser fluence for phase explosion, the pressure and temperature at the target surface and the velocity of the expulsed material were reported. A narrow transition zone was realized between the normal vaporization and phase explosion fields. The proof of heterogeneous boiling was given with detailed micrographs. A novel thermal model was proposed for laser-induced splashing at high fluences. Packaging factor and scattering arc radius terms were proposed to state the level of the melt ejection process. Results of the present investigation explain the explosive boiling during high-power laser interaction with metal.

  18. How long does it take to boil an egg? Revisited

    Energy Technology Data Exchange (ETDEWEB)

    Buay, D [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Foong, S K [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Kiang, D [Department of Physics, Chinese University of Hong Kong, Shatin, New Territories, Hong Kong (China); Kuppan, L [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Centre for Research in Pedagogy and Practice, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Liew, V H [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore)

    2006-01-01

    How long does it take to boil an egg? Theoretical prediction, based on a simple adaptation of the solution to the exact thermal diffusion equation for a sphere, is consistent with experiments. The experimental data are also used to estimate an average value for the thermal diffusivity of an egg.

  19. 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 - appendices. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

    The NRC staff is in need of decommissioning 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) 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 presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clear 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.

  20. Melt Dispersion and Direct Containment Heating (DCH) Experiments für KONVOI reactors (KIT Scientific Reports ; 7567)

    OpenAIRE

    Meyer, Leonhard

    2010-01-01

    The DISCO-H test facility was used to perform scaled experiments that simulate melt ejection scenarios under low system pressure in Severe Accidents in Pressurized Water Reactors (PWR). These experiments are designed to investigate the fluid-dynamic, thermal and chemical processes during melt ejection out of a breach in the lower head of a PWR pressure vessel at pressures around and below 2 MPa with an iron-alumina melt and steam. The report presents results from a test series with the geomet...

  1. The Explosive Boiling Experiment Study on Underwater Laser Processing%水下激光加工的爆发沸腾实验研究

    Institute of Scientific and Technical Information of China (English)

    2016-01-01

    In order to find an appropriate focus position of underwater pulse laser processing ,and then compared with the processing effect in the air ,the experiment by changed the laser focal distance to bombardment thermal conductive silicone piece was carried out. Used a laser scanning confocal microscope to measure the pit size ,and draw the pit depth/width-focus distance trend chart. Used a high-speed photography to shoot underwater laser processing phenomenon ,got the phenomenon of plasma flash and cavitation bubbles ,this phenomenon had been theoretical analysis and experimental verification. The results show that when the focus position is below the surface of the material , mechanism of underwater laser processing is heat transfer ,the plasma shock wave and cavitation bubble burst combination result. Laser machining efficiency is closely related with the laser focal position.%为了探寻水下脉冲激光加工的合适焦点位置,并与空气中的加工效果进行对比,采用改变激光焦点位置轰击导热硅胶片的方法进行实验.用激光扫描共聚焦显微镜测量凹坑尺寸,并绘制了凹坑深度、宽度与焦点位置趋势图.用高速摄影仪拍摄水下激光加工现象,得到了等离子体闪光及其诱导产生的空化泡现象,并进行了理论分析和实验验证.结果表明:当焦点位置在材料表面以下时,水下激光加工是热传导、等离子体冲击波及空化泡冲击波共同作用的结果,激光加工效率与激光焦点位置密切相关.

  2. Considerations on a critical experiment program for a large fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-09-15

    The design studies for large LMFBR in Japan are being continued by PNC and Electric Companies. Both of them have adopted loop-type and 1000 MWe class reactor as the reference design, but main parameters of the core have not yet been fixed. The main core parameters of the present design are shown. Comparing the geometrical properties with those in ZPPR-9 and 10, some of the degrees of mockup are not satisfactory. In addition, there is another difference between a reactor and a mockup critical assembly. Therefore extrapolation is important to apply the results of JUPITER to the core design.

  3. Design study of a fast spectrum zero-power reactor dedicated to source driven sub-critical experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mercatali, L.; Serikov, A. [Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Baeten, P.; Uyttenhove, W. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lafuente, A. [Univerisdad Politecnica de Madrid, 28006 Madrid (Spain); Teles, P. [Instituto Tecnologico e Nuclear, EN 10, 2680-953 Sacavem (Portugal)

    2010-09-15

    In the framework of the European P and T program (IFP6-EUROTRANS), the Generation of Uninterrupted Intense NEutrons pulses at the lead VEnus REactor (GUINEVERE) project consists of an Accelerator Driven System (ADS) that is composed by a fast lead simulated-cooled reactor operated in sub-critical conditions, coupled with an updated version of the GENEPI neutron generator previously used for the MUSE experiments. The GUINEVERE facility aims at developing and improving different techniques for the reactivity monitoring of sub-critical ADS's. As such, the GUINEVERE project will comprise a series of major experiments that will be performed in the near future. The GUINEVERE facility will be located at the VENUS light water moderated research reactor at the SCK-CEN site of Mol (Belgium), which needs to be modified in order to accommodate a completely different and new type of core. A series of constraints were taken into account in the technical design of the GUINEVERE core, in order to properly conjugate the technical feasibility of this facility and the necessity to comply with the envisioned experimental program and its associated scientific outcome. The complete design study of the GUINEVERE core is the subject of this paper. The final design of the fuel assemblies, safety and control rods is provided. Also, the critical core configuration, to be used as reference for absolute reactivity measurements, is presented along with its associated reactor physics parameters, calculated by means of Monte Carlo methodologies. Finally, for licensing purposes, the GUINEVERE facility must satisfy the required nuclear safety criteria of the Belgian safety authorities, and in this paper, an overview of the safety analysis that has been performed with regard to the core physics, thermal assessment and shielding issues is also provided. (author)

  4. Reactor Bolshoi Moshchnosti Kalani; Reacteurs RBMK

    Energy Technology Data Exchange (ETDEWEB)

    Bastien, D. [Conservatoire National des Arts et Metiers (CNAM), 75 - Paris (France)

    2000-01-01

    The Reactor Bolshoi Molshchnosti Kalani (RBMK) are pressure tubes reactor, boiling light water cooled. Exported since 1990 from the ex-USSR, they are today in three independent countries: Russian, Ukraine and Lithuania. Since this date, data exchange with the occident allowed the better knowledge of this reactor type. The design, the technical description (core, fuel, primary system), the safety and the improvement since Chernobyl are detailed. (A.L.B.)

  5. Pool Boiling of FC 770 on Graphene Oxide Coatings: A Study of Critical Heat Flux and Boiling Heat Transfer Enhancement Mechanisms

    OpenAIRE

    Sayee Mohan, Kaushik

    2016-01-01

    This thesis investigates pool boiling heat transfer from bare and graphene-coated NiCr wires in a saturated liquid of FC 770, a fluorocarbon fluid. Of particular interest was the effect of graphene-oxide platelets, dip-coated onto the heater surface, in enhancing the nucleate boiling heat transfer (BHT) rates and the critical heat flux (CHF) value. In the course of the pool boiling experiment, the primary focus was on the reduction mechanism of graphene oxide. The transition from hydrophilic ...

  6. Boiling heat transfer on fins – experimental and numerical procedure

    Directory of Open Access Journals (Sweden)

    Orzechowski T.

    2014-03-01

    Full Text Available The paper presents the research methodology, the test facility and the results of investigations into non-isothermal surfaces in water boiling at atmospheric pressure, together with a discussion of errors. The investigations were conducted for two aluminium samples with technically smooth surfaces and thickness of 4 mm and 10 mm, respectively. For the sample of lower thickness, on the basis of the surface temperature distribution measured with an infrared camera, the local heat flux and the heat transfer coefficient were determined and shown in the form of a boiling curve. For the thicker sample, for which 1-D model cannot be used, numerical calculations were conducted. They resulted in obtaining the values of the local heat flux on the surface the invisible to the infrared, camera i.e. on the side on which the boiling of the medium proceeds.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  8. Measurement of neutrinos released in nuclear reactors through the Borexino experiment; Mesure des neutrinos de reacteurs nucleaires dans l'experience Borexino

    Energy Technology Data Exchange (ETDEWEB)

    Dadoun, O

    2003-06-01

    The main goal of the Borexino experiment is to measure in real time the solar neutrino flux from the beryllium (Be{sup 7}) line at 862 keV. Beyond this pioneer low energy neutrino detection, Borexino will be able to measure solar neutrinos above the MeV, (B{sup 8} neutrinos and pep neutrinos), nuclear reactor neutrinos (with an average energy of 3 MeV) and the supernova neutrinos (their spectrum goes up to some ten MeV). In this work I mainly focus on the study of the nuclear reactors neutrinos. This field has recently been enriched by the results of the KamLAND experiment, which have greatly improved the determination of the neutrino oscillation parameters. In order to measure these events which are above the MeV, the Borexino collaboration entrusted the PCC group at College de France, with the tasks of developing a fast digit system running at 400 MHz: the FADC cards. The PCC group designed the FADC cards and completed them at the beginning of 2002. The first cards which were introduced in the main electronic acquisition unit allowed us to control their functioning and that of the acquisition software. FADC cards were also installed in the Borexino prototype, CTF. The data are analysed in order to determine a limit to the expected background noise of Borexino in measuring the nuclear reactor neutrinos. (author)

  9. Optimization of Quick-boiling Process for Lotus Root Slices%藕片水焯工艺优化

    Institute of Scientific and Technical Information of China (English)

    计红芳; 张令文; 杨铭铎; 李光磊; 蔡振林

    2011-01-01

    Lotus root slices was quick boiled in four different media such as water,acetic acid,sodium chloride and sodium bicarbonate solution and as a result,acetic acid was identified as the best medium through comprehensive considering VC content and sensory score of quick-boiled lotus root slices as well as dry matter dissolution rate.Further,on the basis of one-factor-at-a-time experiments,Box-Behnken design combined with response surface methodology was used to optimize process conditions for the quick boiling of lotus root slices in acetic acid.The results indicated that the optimal quick boiling conditions were acetic acid concentration of 0.35%,temperature of 68.4 ℃ and quick-boiling time of 2.4 min.After quick boiling under the optimal process conditions,the content of vitamin C in lotus root slices was 30.185 mg/100 g fresh weight,and better eating quality was also achieved.The results of variance analysis showed that acetic acid concentration,quick-boiling temperature and quick-boiling time had significant effects on the content of vitamin C in quick-boiled lotus root slices(P0.01).Moreover,the cross-interactions between acetic acid concentration and temperature and between temperature and quick-boiling time also revealed a significant effect on the content of vitamin C in lotus root slices(P0.05).%以新鲜莲藕为试验材料,对水、乙酸溶液、食盐溶液和碳酸氢钠溶液等4种水焯介质进行探讨;通过单因素试验和Box-Behnken设计及响应面分析法,对藕片水焯的最佳工艺条件进行研究。结果表明:乙酸可作为藕片的水焯介质;藕片水焯的最佳工艺条件为乙酸体积分数0.35%、温度68.4℃、时间2.4min,在该条件下藕片中VC含量为30.185mg/100g,干物质溶出率为5.384%,藕片食用品质可达到较高水平。响应面方差分析结果显示,乙酸体积分数、温度和水焯时间对水焯藕片VC含量有显著影响(P〈0.01),乙酸体积分数和温度、温度和

  10. 9 CFR 75.4 - Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic...

    Science.gov (United States)

    2010-01-01

    ... IN HORSES, ASSES, PONIES, MULES, AND ZEBRAS Equine Infectious Anemia (swamp Fever) § 75.4 Interstate..., including name, age, sex, breed, color, and markings. Reactor. Any horse, ass, mule, pony or zebra which is... to prevent the transmission of equine infectious anemia to other horses, asses, ponies, mules,...

  11. Reactivity worth measurements on fast burst reactor Caliban - description and interpretation of integral experiments for the validation of nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Richard, B. [Commissariat a l' Energie Atomique et Aux Energies Alternatives CEA, DAM, VALDUC, F-21120 Is-sur-Tille (France)

    2012-07-01

    Reactivity perturbation experiments using various materials are being performed on the HEU fast core CALIBAN, an experimental device operated by the CEA VALDUC Criticality and Neutron Transport Res