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Sample records for melting model bwr

  1. BWR core melt progression phenomena: Experimental analyses

    International Nuclear Information System (INIS)

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

  2. Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

    International Nuclear Information System (INIS)

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-01-01

    Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and

  3. Final results of the XR2-1 BWR metallic melt relocation experiment

    International Nuclear Information System (INIS)

    Gauntt, R.O.; Humphries, L.L.

    1997-08-01

    This report documents the final results of the XR2-1 boiling water reactor (BWR) metallic melt relocation experiment, conducted at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The objective of this experiment was to investigate the material relocation processes and relocation pathways in a dry BWR core following a severe nuclear reactor accident such as an unrecovered station blackout accident. The imposed test conditions (initial thermal state and the melt generation rates) simulated the conditions for the postulated accident scenario and the prototypic design of the lower core test section (in composition and in geometry) ensured that thermal masses and physical flow barriers were modeled adequately. The experiment has shown that, under dry core conditions, the metallic core materials that melt and drain from the upper core regions can drain from the core region entirely without formation of robust coherent blockages in the lower core. Temporary blockages that suspended pools of molten metal later melted, allowing the metals to continue draining downward. The test facility and instrumentation are described in detail. The test progression and results are presented and compared to MERIS code analyses. 6 refs., 55 figs., 4 tabs

  4. BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

    International Nuclear Information System (INIS)

    Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

  5. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  6. A simplified spatial model for BWR stability

    International Nuclear Information System (INIS)

    Berman, Y.; Lederer, Y.; Meron, E.

    2012-01-01

    A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

  7. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  8. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

    Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

    1990-01-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  9. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

    Ballestrin Bolea, J.M.; Blazquez, J.B.

    1990-01-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical struct-ure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations in non-linear. Simple parametric calculat-ion of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author). 7 refs

  10. Model of interfacial melting

    DEFF Research Database (Denmark)

    Mouritsen, Ole G.; Zuckermann, Martin J.

    1987-01-01

    A two-dimensional model is proposed to describe systems with phase transitions which take place in terms of crystalline as well as internal degrees of freedom. Computer simulation of the model shows that the interplay between the two sets of degrees of freedom permits observation of grain-boundar......-boundary formation and interfacial melting, a nonequilibrium process by which the system melts at the boundaries of a polycrystalline domain structure. Lipid membranes are candidates for systems with pronounced interfacial melting behavior....

  11. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  12. Physical model of nonlinear noise with application to BWR stability

    International Nuclear Information System (INIS)

    March-Leuba, J.; Perez, R.B.

    1983-01-01

    Within the framework of the present model it is shown that the BWR reactor cannot be unstable in the linear sense, but rather it executes limited power oscillations of a magnitude that depends on the operating conditions. The onset of these oscillations can be diagnosed by the decrease in stochasticity in the power traces and by the appearance of harmonics in the PSD

  13. BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

    Energy Technology Data Exchange (ETDEWEB)

    Ballestrin Bolea, J M; Blazquez Martinez, J B

    1990-07-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  14. Initiation model for intergranular stress corrosion cracking in BWR pipes

    International Nuclear Information System (INIS)

    Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

    1981-01-01

    Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

  15. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    International Nuclear Information System (INIS)

    Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.

    2008-01-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  16. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)

    2008-07-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  17. A detailed BWR recirculation loop model for RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Araiza-Martínez, Enrique, E-mail: enrique.araiza@inin.gob.mx; Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx; Castillo-Durán, Rogelio, E-mail: rogelio.castillo@inin.gob.mx

    2017-01-15

    Highlights: • A new detailed BWR recirculation loop model was developed for RELAP. • All jet pumps, risers, manifold, suction and control valves, and recirculation pump are modeled. • Model is tested against data from partial blockage of two jet pumps. • For practical applications, simulation results showed good agreement with available data. - Abstract: A new detailed geometric model of the whole recirculation loop of a BWR has been developed for the code RELAP. This detailed model includes the 10 jet pumps, 5 risers, manifold, suction and control valves, and the recirculation pump, per recirculation loop. The model is tested against data from an event of partial blockage at the entrance nozzle of one jet pump in both recirculation loops. For practical applications, simulation results showed good agreement with data. Then, values of parameters considered as figure of merit (reactor power, dome pressure, core flow, among others) for this event are compared against those from the common 1 jet pump per loop model. The results show that new detailed model led to a closer prediction of the reported power change. The detailed recirculation loop model can provide more reliable boundary condition data to a CFD models for studies of, for example, flow induced vibration, wear, and crack initiation.

  18. Studies on core melt behaviour in a BWR pressure vessel lower head

    International Nuclear Information System (INIS)

    Lindholm, I.; Ikonen, K.; Hedberg, K.

    1999-01-01

    Core debris behaviour in the Nordic BWR lower head was investigated numerically using MELCOR and MAAP4 codes. Lower head failure due to penetration failure was studied with more detailed PASULA code taking thermal boundary conditions from MELCOR calculations. Creep rupture failure mode was examined with the two integral codes. Also, the possibility to prevent vessel failure by late reflooding was assessed in this study. (authors)

  19. Logical model for the control of a BWR turbine

    International Nuclear Information System (INIS)

    Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.

    2009-01-01

    In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

  20. System control model of a turbine for a BWR

    International Nuclear Information System (INIS)

    Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A.

    2009-10-01

    In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

  1. Development of a parametric containment event tree model for a severe BWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T [OTO-Consulting Ay, Helsinki (Finland)

    1995-04-01

    A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

  2. Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

    International Nuclear Information System (INIS)

    Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

    2002-01-01

    Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

  3. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

  4. Simplified distributed parameters BWR dynamic model for transient and stability analysis

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

    2006-01-01

    This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

  5. Modeling of SCC initiation and propagation mechanisms in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmeister, Hans, E-mail: Hans.Hoffmeister@hsu-hh.de [Institute for Failure Analysis and Failure Prevention ISSV e.V., c/o Helmut Schmidt University of the Federal Armed Forces, D-22039 Hamburg (Germany); Klein, Oliver [Institute for Failure Analysis and Failure Prevention ISSV e.V., c/o Helmut Schmidt University of the Federal Armed Forces, D-22039 Hamburg (Germany)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer We show that SSC in BWR environments includes anodic crack propagation and hydrogen assisted cracking. Black-Right-Pointing-Pointer Hydrogen cracking is triggered by crack tip acidification following local impurity accumulations and subsequent phase precipitations. Black-Right-Pointing-Pointer We calculate effects of pH, chlorides, potentials and stress on crack SCC growth rates at 288 Degree-Sign C. - Abstract: During operation of mainly BWRs' (Boiling Water Reactors) excursions from recommended water chemistries may provide favorite conditions for stress corrosion cracking (SCC). Maximum levels for chloride and sulfate ion contents for avoiding local corrosion are therefore given in respective water specifications. In a previously published deterministic 288 Degree-Sign C - corrosion model for Nickel as a main alloying element of BWR components it was demonstrated that, as a theoretically worst case, bulk water chloride levels as low as 30 ppb provide local chloride ion accumulation, dissolution of passivating nickel oxide and precipitation of nickel chlorides followed by subsequent local acidification. In an extension of the above model to SCC the following work shows that, in a first step, local anodic path corrosion with subsequent oxide breakdown, chloride salt formation and acidification at 288 Degree-Sign C would establish local cathodic reduction of accumulated hydrogen ions inside the crack tip fluid. In a second step, local hydrogen reduction charges and increasing local crack tip strains from increasing crack lengths at given global stresses are time stepwise calculated and related to experimentally determined crack critical cathodic hydrogen charges and fracture strains taken from small scale SSRT tensile tests pieces. As a result, at local hydrogen equilibrium potentials higher than those of nickel in the crack tip solution, hydrogen ion reduction initiates hydrogen crack propagation that is enhanced with

  6. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    1999-05-01

    This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

  7. GOTHIC MODEL OF BWR SECONDARY CONTAINMENT DRAWDOWN ANALYSES

    International Nuclear Information System (INIS)

    Hansen, P.N.

    2004-01-01

    This article introduces a GOTHIC version 7.1 model of the Secondary Containment Reactor Building Post LOCA drawdown analysis for a BWR. GOTHIC is an EPRI sponsored thermal hydraulic code. This analysis is required by the Utility to demonstrate an ability to restore and maintain the Secondary Containment Reactor Building negative pressure condition. The technical and regulatory issues associated with this modeling are presented. The analysis includes the affect of wind, elevation and thermal impacts on pressure conditions. The model includes a multiple volume representation which includes the spent fuel pool. In addition, heat sources and sinks are modeled as one dimensional heat conductors. The leakage into the building is modeled to include both laminar as well as turbulent behavior as established by actual plant test data. The GOTHIC code provides components to model heat exchangers used to provide fuel pool cooling as well as area cooling via air coolers. The results of the evaluation are used to demonstrate the time that the Reactor Building is at a pressure that exceeds external conditions. This time period is established with the GOTHIC model based on the worst case pressure conditions on the building. For this time period the Utility must assume the primary containment leakage goes directly to the environment. Once the building pressure is restored below outside conditions the release to the environment can be credited as a filtered release

  8. Investigation of the Potential for In-Vessel Melt Retention in the Lower Head of a BWR by Cooling through the Control Rod Guide Tubes. APRl 4, Stage 2 Report

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Jasiulevicius, A.; Konovalikhin, M.

    2004-01-01

    recommended that further investigations, both experimental and model development, be conducted to (a) check reproducibility of data (b) employ different flow rates (c) employ different simulant materials and (d) develop a comprehensive model, in order to certify that the coolability that can be achieved with establishing a water flow in the CRGTs will be able to retain the melt in the lower head of a BWR. We believe it will be an extremely important accident management strategy for a Swedish BWR since it will obviate the consideration of the prime licensing issue of ex-vessel steam explosion induced containment failure associated with the present scheme of establishing a water pool in the lower drywell of all the Swedish BWRs

  9. System control model of a turbine for a BWR; Modelo del sistema de control de una turbina para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rodolfo.amador@inin.gob.mx

    2009-10-15

    In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

  10. Identification of the reduced order models of a BWR reactor

    International Nuclear Information System (INIS)

    Hernandez S, A.

    2004-01-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  11. Identification of BWR feedwater control system using autoregressive integrated model

    International Nuclear Information System (INIS)

    Kanemoto, Shigeru; Andoh, Yasumasa; Yamamoto, Fumiaki; Idesawa, Masato; Itoh, Kazuo.

    1983-01-01

    With the view of contributing toward more reliable interpretation of noise behavior under normal operating conditions, which is essential for correct detection and/or diagnosis of incipient anomalies in nuclear power plants by noise analysis technique, studies has been undertaken of the noise behavior in a BWR feedwater control system, with use made of a multivariate autoregressive modeling technique. Noise propagation mechanisms as well as open- and closed-loop responses in the system are identified from noise data by a method in which an autoregressive integrated model is introduced. The closed-loop responses obtained with this method are compared with transient data from an actual test, and confirmed to be reliable in estimating semi-quantitative features. Other analyses performed with this model also yield results that appear most reasonable in their physical characteristics. These results have demonstrated the effectiveness of the noise analyses technique based on the autoregressive integrated model for evaluating and diagnosing the performance of feedwater control systems. (author)

  12. BWR regional instability model and verification on ringhals-1 test

    International Nuclear Information System (INIS)

    Hotta, Akitoshi; Suzawa, Yojiro

    1996-01-01

    Regional instability is known as one type of the coupled neutronic-thermohydraulic phenomena of boiling water reactors (BWRs), where the thermohydraulic density wave propagation mechanism is predominant. Historically, it has been simulated by the three-dimensional time domain code in spite of its significant computing time. On the other hand, there have been proposals to apply the frequency domain models in regional instability considering the subcriticality of the higher neutronic mode. However, their application still remains in corewide instability mainly because of the lack of more detailed methodological and empirical studies. In this study, the current version of the frequency domain model was extended and verified based on actual core regional instability measurement data. The mathematical model LAPUR, the well-known frequency domain stability code, was reviewed from the standpoint of pure thermohydraulics and neutronic-thermohydraulic interaction mechanisms. Based on the ex-core loop test data, the original LAPUR mixed friction and local pressure loss model was modified, taking into account the different dynamic behavior of these two pressure-loss mechanisms. The perturbation term of the two-phase friction multiplier, which is the sum of the derivative of void fraction and subcool enthalpy, was adjusted theoretically. The adequacy of the instability evaluation system was verified based on the Ringhals unit 1 test data, which were supplied to participants of the Organization for Economic Cooperation and Development/Nuclear Energy Agency BWR Stability Benchmark Project

  13. Results of modeling advanced BWR fuel designs using CASMO-4

    International Nuclear Information System (INIS)

    Knott, D.; Edenius, M.

    1996-01-01

    Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

  14. Multiscale Models of Melting Arctic Sea Ice

    Science.gov (United States)

    2014-09-30

    Sea ice reflectance or albedo , a key parameter in climate modeling, is primarily determined by melt pond and ice floe configurations. Ice - albedo ...determine their albedo - a key parameter in climate modeling. Here we explore the possibility of a conceptual sea ice climate model passing through a...bifurcation points. Ising model for melt ponds on Arctic sea ice Y. Ma, I. Sudakov, and K. M. Golden Abstract: The albedo of melting

  15. Identification of the reduced order models of a BWR reactor; Identificacion de modelos de orden reducido de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez S, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: augusto@correo.unam.mx

    2004-07-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  16. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  17. Multiscale approach to equilibrating model polymer melts

    DEFF Research Database (Denmark)

    Svaneborg, Carsten; Ali Karimi-Varzaneh, Hossein; Hojdis, Nils

    2016-01-01

    We present an effective and simple multiscale method for equilibrating Kremer Grest model polymer melts of varying stiffness. In our approach, we progressively equilibrate the melt structure above the tube scale, inside the tube and finally at the monomeric scale. We make use of models designed...

  18. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

    2005-01-01

    Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

  19. Fission product model for BWR analysis with improved accuracy in high burnup

    International Nuclear Information System (INIS)

    Ikehara, Tadashi; Yamamoto, Munenari; Ando, Yoshihira

    1998-01-01

    A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1%Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation. (author)

  20. Improved point-kinetics model for the BWR control rod drop accident

    International Nuclear Information System (INIS)

    Neogy, P.; Wakabayashi, T.; Carew, J.F.

    1985-01-01

    A simple prescription to account for spatial feedback weighting effects in RDA (rod drop accident) point-kinetics analyses has been derived and tested. The point-kinetics feedback model is linear in the core peaking factor, F/sub Q/, and in the core average void fraction and fuel temperature. Comparison with detailed spatial kinetics analyses indicates that the improved point-kinetics model provides an accurate description of the BWR RDA

  1. Development of an Input Model to MELCOR 1.8.5 for the Oskarshamn 3 BWR

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Lars [Lentek, Nykoeping (Sweden)

    2006-05-15

    An input model has been prepared to the code MELCOR 1.8.5 for the Swedish Oskarshamn 3 Boiling Water Reactor (O3). This report describes the modelling work and the various files which comprise the input deck. Input data are mainly based on original drawings and system descriptions made available by courtesy of OKG AB. Comparison and check of some primary system data were made against an O3 input file to the SCDAP/RELAP5 code that was used in the SARA project. Useful information was also obtained from the FSAR (Final Safety Analysis Report) for O3 and the SKI report '2003 Stoerningshandboken BWR'. The input models the O3 reactor at its current state with the operating power of 3300 MW{sub th}. One aim with this work is that the MELCOR input could also be used for power upgrading studies. All fuel assemblies are thus assumed to consist of the new Westinghouse-Atom's SVEA-96 Optima2 fuel. MELCOR is a severe accident code developed by Sandia National Laboratory under contract from the U.S. Nuclear Regulatory Commission (NRC). MELCOR is a successor to STCP (Source Term Code Package) and has thus a long evolutionary history. The input described here is adapted to the latest version 1.8.5 available when the work began. It was released the year 2000, but a new version 1.8.6 was distributed recently. Conversion to the new version is recommended. (During the writing of this report still another code version, MELCOR 2.0, has been announced to be released within short.) In version 1.8.5 there is an option to describe the accident progression in the lower plenum and the melt-through of the reactor vessel bottom in more detail by use of the Bottom Head (BH) package developed by Oak Ridge National Laboratory especially for BWRs. This is in addition to the ordinary MELCOR COR package. Since problems arose running with the BH input two versions of the O3 input deck were produced, a NONBH and a BH deck. The BH package is no longer a separate package in the new 1

  2. Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments; TOPICAL

    International Nuclear Information System (INIS)

    Ott, L.J.

    1994-01-01

    The objectives of the Boiling Water Reactor Experimental Analysis and Model Development for Severe Accidents (BEAMD) Program at the Oak Ridge National Laboratory (ORNL) are: (1) the development of a sound quantitative understanding of boiling water reactor (BWR) core melt progression; this includes control blade and channel box effects, metallic melt relocation and possible blockage formation under severe accident conditions, and (2) provision of BWR melt progression modeling capabilities in SCDAP/RELAP5 (consistent with the BWR experimental data base). This requires the assessment of current modeling of BWR core melt progression against the expanding BWR data base. Emphasis is placed upon data from the BWR tests in the German CORA test facility and from the ex-reactor experiments[Sandia National Laboratories (SNL)] on metallic melt relocation and blockage formation in BWRs, as well as upon in-reactor data from the Annular Core Research Reactor (ACRR) DF-4 BWR test (conducted in 1986 at SNL). The BEAMD Program is a derivative of the BWR Severe Accident Technology Programs at ORNL. The ORNL BWR programs have studied postulated severe accidents in BWRs and have developed a set of models specific to boiling water reactor response under severe accident conditions. These models, in an experiment-specific format, have been successfully applied to both pretest and posttest analyses of the DF-4 experiment, and the BWR severe fuel damage (SFD) experiments performed in the CORA facility at the Kernforschungszentrum Karlsruhe (KfK) in Germany, resulting in excellent agreement between model prediction and experiment. The ORNL BWR models have provided for more precise predictions of the conditions in the BWR experiments than were previously available. This has provided a basis for more accurate interpretation of the phenomena for which the experiments are performed. The experiment-specific models, as used in the ORNL DF-4 and CORA BWR experimental analyses, also provide a basis

  3. Modeling the summertime evolution of sea-ice melt ponds

    DEFF Research Database (Denmark)

    Lüthje, Mikael; Feltham, D.L.; Taylor, P.D.

    2006-01-01

    We present a mathematical model describing the summer melting of sea ice. We simulate the evolution of melt ponds and determine area coverage and total surface ablation. The model predictions are tested for sensitivity to the melt rate of unponded ice, enhanced melt rate beneath the melt ponds...

  4. Description of the power plant model BWR-plasim outlined for the Barsebaeck 2 plant

    International Nuclear Information System (INIS)

    Christensen, P. la Cour.

    1979-08-01

    A description is given of a BWR power plant model outlined for the Barsebaeck 2 plant with data placed at our disposal by the Swedish Power Company Sydkraft A/B. The basic operations are derived and simplifications discussed. The model is implemented with a simulation system DYSYS which assures reliable solutions and easy programming. Emphasis has been placed on the models versatility and flexibility so new features are easy to incorporate. The model may be used for transient calculations for both normal plant conditions and for abnormal occurences as well as for control system studies. (author)

  5. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

    2017-03-15

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  6. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    International Nuclear Information System (INIS)

    Badea, Aurelian F.; Cacuci, Dan G.

    2017-01-01

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  7. Study on vertical seismic response model of BWR-type reactor building

    International Nuclear Information System (INIS)

    Konno, T.; Motohashi, S.; Izumi, M.; Iizuka, S.

    1993-01-01

    A study on advanced seismic design for LWR has been carried out by the Nuclear Power Engineering Corporation (NUPEC), under the sponsorship of the Ministry of International Trade and Industry (MITI) of Japan. As a part of the study, it has been investigated to construct an accurate analytical model of reactor buildings for a seismic response analysis, which can reasonably represent dynamic characteristics of the building. In Japan, vibration models of reactor buildings for horizontal ground motion have been studied and examined through many simulation analyses for forced vibration tests and earthquake observations of actual buildings. And now it is possible to establish a reliable horizontal vibration model on the basis of multi-lumped mass and spring model. However, vertical vibration models have not been so much studied as horizontal models, due to less observed data for vertical motions. In this paper, the vertical seismic response models of a BWR-type reactor building including soil-structure interaction effect are numerically studied, by comparing the dynamic characteristics of (1) three dimensional finite element model, (2) multi-stick lumped mass model with a flexible base-mat, (3) multi-stick lumped mass model with a rigid base-mat and (4) single-stick lumped mass model. In particular, the BWR-type reactor building has the long span truss roof which is considered to be one of the critical members to vertical excitation. The modelings of the roof trusses are also studied

  8. Logical model for the control of a BWR turbine;Modelo logico para el control de una turbina de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R., E-mail: yonaeton@hotmail.co [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2009-07-01

    In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

  9. Modelling perspectives on radiation chemistry in BWR reactor core

    International Nuclear Information System (INIS)

    Ibe, Eishi

    1991-01-01

    Development of a full-system boiling water reactor core model started in 1982. The model included a two-region reactor core, one with and one without boiling. Key design parameters consider variable dose rates in a three-layer liquid downcomer. Dose rates in the core and downcomer include both generation and recombination reactions of species. Agreement is good between calculations and experimental data of oxygen concentration as a function of hydrogen concentration for different bubble sizes. Oxygen concentration is reduced in the reactor pressure vessel (RPV) by increasing bubble size. The multilayer model follows the oxygen data better than a single-layered model at high concentrations of hydrogen. Key reactions are reduced to five radiolysis reactions and four decomposition reactions for hydrogen peroxide. Calculations by the DOT 3 code showed dose rates from neutrons and gamma rays in various parts of the core. Concentrations of oxygen, hydrogen peroxide, and hydrogen were calculated by the model as a function of time from core inlet. Similar calculations for NWC and HWC were made as a function of height from core inlet both in the boiling channel an the bypass channel. Finally the model was applied to calculate the oxygen plus half the hydrogen peroxide concentrations as a function of hydrogen concentration to compare with data from five plants. Power density distribution with core height was given for an early stage and an end stage of a cycle. Increases of dose rates in the turbine for seven plants were shown as a function of increased hydrogen concentration in the reactor water

  10. Models and observations of Arctic melt ponds

    Science.gov (United States)

    Golden, K. M.

    2016-12-01

    During the Arctic melt season, the sea ice surface undergoes a striking transformation from vast expanses of snow covered ice to complex mosaics of ice and melt ponds. Sea ice albedo, a key parameter in climate modeling, is largely determined by the complex evolution of melt pond configurations. In fact, ice-albedo feedback has played a significant role in the recent declines of the summer Arctic sea ice pack. However, understanding melt pond evolution remains a challenge to improving climate projections. It has been found that as the ponds grow and coalesce, the fractal dimension of their boundaries undergoes a transition from 1 to about 2, around a critical length scale of 100 square meters in area. As the ponds evolve they take complex, self-similar shapes with boundaries resembling space-filling curves. I will outline how mathematical models of composite materials and statistical physics, such as percolation and Ising models, are being used to describe this evolution and predict key geometrical parameters that agree very closely with observations.

  11. BWR Mark III containment analyses using a GOTHIC 8.0 3D model

    International Nuclear Information System (INIS)

    Jimenez, Gonzalo; Serrano, César; Lopez-Alonso, Emma; Molina, M del Carmen; Calvo, Daniel; García, Javier; Queral, César; Zuriaga, J. Vicente; González, Montserrat

    2015-01-01

    Highlights: • The development of a 3D GOTHIC code model of BWR Mark-III containment is described. • Suppression pool modelling based on the POOLEX STB-20 and STB-16 experimental tests. • LOCA and SBO transient simulated to verify the behaviour of the 3D GOTHIC model. • Comparison between the 3D GOTHIC model and MAAP4.07 model is conducted. • Accurate reproduction of pre severe accident conditions with the 3D GOTHIC model. - Abstract: The purpose of this study is to establish a detailed three-dimensional model of Cofrentes NPP BWR/6 Mark III containment building using the containment code GOTHIC 8.0. This paper presents the model construction, the phenomenology tests conducted and the selected transient for the model evaluation. In order to study the proper settings for the model in the suppression pool, two experiments conducted with the experimental installation POOLEX have been simulated, allowing to obtain a proper behaviour of the model under different suppression pool phenomenology. In the transient analyses, a Loss of Coolant Accident (LOCA) and a Station Blackout (SBO) transient have been performed. The main results of the simulations of those transients were qualitative compared with the results obtained from simulations with MAAP 4.07 Cofrentes NPP model, used by the plant for simulating severe accidents. From this comparison, a verification of the model in terms of pressurization, asymmetric discharges and high pressure release were obtained. The completeness of this model has proved to adequately simulate the thermal hydraulic phenomena which occur in the containment during accidental sequences

  12. Analytical modeling of bwr safety relief valve blowdown phenomenon

    International Nuclear Information System (INIS)

    Hwang, J.G.; Singh, A.

    1984-01-01

    An analytical, qualitative understanding of the pool pressures measured during safety relief valve discharge in boiling water reactors equipped with X-quenchers has been developed and compared to experimental data. A pressure trace typically consists of a brief 25-35 Hz. oscillation followed by longer 5-15 Hz. oscillation. In order to explain the pressure response, a discharge line vent clearing model has been coupled with a Rayleigh bubble dynamic model. The local conditions inside the safety relief valve discharge lines and inside of the X-quencher were simulated successfully with RELAP5. The simulation allows one to associate the peak pressure inside the quencher arm with the onset of air discharge into the suppression pool. Using the pressure and thermodynamic quality at quencher exit of RELAP5 calculation as input, a Rayleigh model of pool bubble dynamics has successfully explained both the higher and lower frequency pressure oscillations. The higher frequency oscillations are characteristic of an air bubble emanating from a single row of quencher holes. The lower frequency pressure oscillations are characteristic of a larger air bubble containing all the air expelled from one side of an X-quencher arm

  13. Mathematical model of melt flow channel granulator

    Directory of Open Access Journals (Sweden)

    A. A. Kiselev

    2016-01-01

    Full Text Available Granulation of carbohydrate-vitamin-mineral supplements based on molasses is performed at a high humidity (26 %, so for a stable operation of granulator it is necessary to reveal its melt flow pattern. To describe melt non-isothermal flow in the granulator a mathematical model with following initial equations: continuity equation, motion equation and rheological equation – was developed. The following assumptions were adopted: the melt flow in the granulator is a steady laminar flow; inertial and gravity forces can be ignored; melt is an incompressible fluid; velocity gradient in the flow direction is much smaller than in the transverse direction; the pressure gradient over the cross section of the channel is constant; the flow is hydrodynamically fully developed; effects impact on the channel inlet and outlet may be neglected. Due to the assumptions adopted, it can be considered that in this granulator only velocity components in the x-direction are significant and all the members of the equation with the components and their derivatives with respect to the coordinates y and z can be neglected. The resulting solutions were obtained: the equation for the mean velocity, the equation for determining the volume flow, the formula for calculating of mean time of the melt being in the granulator, the equation for determining the shear stress, the equation for determining the shear rate and the equation for determining the pressure loss. The results of calculations of the equations obtained are in complete agreement with the experimental data; deviation range is 16–19 %. The findings about the melt movement pattern in granulator allowed developing a methodology for calculating a rational design of the granulator molding unit.

  14. A macroscopic cross-section model for BWR pin-by-pin core analysis

    International Nuclear Information System (INIS)

    Fujita, Tatsuya; Endo, Tomohiro; Yamamoto, Akio

    2014-01-01

    A macroscopic cross-section model used in boiling water reactor (BWR) pin-by-pin core analysis is studied. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of core state and depletion history variables and are tabulated prior to core calculations. Variations of cross sections in a core simulator are caused by two different phenomena (i.e. instantaneous and history effects). We treat them through the core state variables and the exposure-averaged core state variables, respectively. Furthermore, the cross-term effect among the core state and the depletion history variables is considered. In order to confirm the calculation accuracy and discuss the treatment of the cross-term effect, the k-infinity and the pin-by-pin fission rate distributions in a single fuel assembly geometry are compared. Some cross-term effects could be negligible since the impacts of them are sufficiently small. However, the cross-term effects among the control rod history (or the void history) and other variables have large impacts; thus, the consideration of them is crucial. The present macroscopic cross-section model, which considers such dominant cross-term effects, well reproduces the reference results and can be a candidate in practical applications for BWR pin-by-pin core analysis on the normal operations. (author)

  15. Development of CFD analysis method based on droplet tracking model for BWR fuel assemblies

    International Nuclear Information System (INIS)

    Onishi, Yoichi; Minato, Akihiko; Ichikawa, Ryoko; Mashara, Yasuhiro

    2011-01-01

    It is well known that the minimum critical power ratio (MCPR) of the boiling water reactor (BWR) fuel assembly depends on the spacer grid type. Recently, improvement of the critical power is being studied by using a spacer grid with mixing devices attaching various types of flow deflectors. In order to predict the critical power of the improved BWR fuel assembly, we have developed an analysis method based on the consideration of detailed thermal-hydraulic mechanism of annular mist flow regime in the subchannels for an arbitrary spacer type. The proposed method is based on a computational fluid dynamics (CFD) model with a droplet tracking model for analyzing the vapor-phase turbulent flow in which droplets are transported in the subchannels of the BWR fuel assembly. We adopted the general-purpose CFD software Advance/FrontFlow/red (AFFr) as the base code, which is a commercial software package created as a part of Japanese national project. AFFr employs a three-dimensional (3D) unstructured grid system for application to complex geometries. First, AFFr was applied to single-phase flows of gas in the present paper. The calculated results were compared with experiments using a round cellular spacer in one subchannel to investigate the influence of the choice of turbulence model. The analyses using the large eddy simulation (LES) and re-normalisation group (RNG) k-ε models were carried out. The results of both the LES and RNG k-ε models show that calculations of velocity distribution and velocity fluctuation distribution in the spacer downstream reproduce the experimental results qualitatively. However, the velocity distribution analyzed by the LES model is better than that by the RNG k-ε model. The velocity fluctuation near the fuel rod, which is important for droplet deposition to the rod, is also simulated well by the LES model. Then, to examine the effect of the spacer shape on the analytical result, the gas flow analyses with the RNG k-ε model were performed

  16. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  17. Melt migration modeling in partially molten upper mantle

    Science.gov (United States)

    Ghods, Abdolreza

    The objective of this thesis is to investigate the importance of melt migration in shaping major characteristics of geological features associated with the partial melting of the upper mantle, such as sea-floor spreading, continental flood basalts and rifting. The partial melting produces permeable partially molten rocks and a buoyant low viscosity melt. Melt migrates through the partially molten rocks, and transfers mass and heat. Due to its much faster velocity and appreciable buoyancy, melt migration has the potential to modify dynamics of the upwelling partially molten plumes. I develop a 2-D, two-phase flow model and apply it to investigate effects of melt migration on the dynamics and melt generation of upwelling mantle plumes and focusing of melt migration beneath mid-ocean ridges. Melt migration changes distribution of the melt-retention buoyancy force and therefore affects the dynamics of the upwelling plume. This is investigated by modeling a plume with a constant initial melt of 10% where no further melting is considered. Melt migration polarizes melt-retention buoyancy force into high and low melt fraction regions at the top and bottom portions of the plume and therefore results in formation of a more slender and faster upwelling plume. Allowing the plume to melt as it ascends through the upper mantle also produces a slender and faster plume. It is shown that melt produced by decompressional melting of the plume migrates to the upper horizons of the plume, increases the upwelling velocity and thus, the volume of melt generated by the plume. Melt migration produces a plume which lacks the mushroom shape observed for the plume models without melt migration. Melt migration forms a high melt fraction layer beneath the sloping base of the impermeable oceanic lithosphere. Using realistic conditions of melting, freezing and melt extraction, I examine whether the high melt fraction layer is able to focus melt from a wide partial melting zone to a narrow region

  18. Modal-based reduced-order model of BWR out-of phase instabilities

    International Nuclear Information System (INIS)

    Turso, J.A.; Edwards, R.M.; March-Leuba, J.

    1995-01-01

    For the past 40 yr, reduced-order modeling of boiling water reactor (BWR) dynamic behavior has been accomplished by several researchers. These models have been primarily concerned with providing insight into the so-called corewide neutron flux oscillation, where the power at each radial location in the core oscillates in unison. This is generally considered to be an illustration of the fundamental neutronic mode excited by the core thermal hydraulics. The time dependence of the fundamental mode is typically described by the point-kinetics equations, with one or more delayed-neutron groups. Thermal-hydraulic excitation of the first azimuthal harmonic mode, the so-called out-of-phase (OOP) instability, has been observed in operating BWRs. The temporal behavior of a low-order model of this phenomenon can be characterized using the modal point-kinetics formulation developed in this paper

  19. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; Bjerke, M.A.; Morrison, G.W.; Petrie, L.M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given

  20. Modelling of the controlled melt flow in a glass melting space – Its melting performance and heat losses

    Czech Academy of Sciences Publication Activity Database

    Jebavá, Marcela; Dyrčíková, Petra; Němec, Lubomír

    2015-01-01

    Roč. 430, DEC 15 (2015), s. 52-63 ISSN 0022-3093 Institutional support: RVO:67985891 Keywords : glass melt flow * mathematical modelling * energy distribution * space utilizatios * melting performance Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass Impact factor: 1.825, year: 2015

  1. General model for Pc-based simulation of PWR and BWR plant components

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, W M; Abomustafa, A M [Faculty of enginnering, alfateh univerity Tripoli, (Libyan Arab Jamahiriya)

    1995-10-01

    In this paper, we present a basic mathematical model derived from physical principles to suit the simulation of PWR-components such as pressurizer, intact steam generator, ruptured steam generator, and the reactor component of a BWR-plant. In our development, we produced an NMMS-package for nuclear modular modelling simulation. Such package is installed on a personal computer and it is designed to be user friendly through color graphics windows interfacing. The package works under three environments, namely, pre-processor, simulation, and post-processor. Our analysis of results using cross graphing technique for steam generator tube rupture (SGTR) accident, yielded a new proposal for on-line monitoring of control strategy of SGTR-accident for nuclear or conventional power plant. 4 figs.

  2. The corrosion potential of stainless steel in BWR environment comparison of data and modeling results

    International Nuclear Information System (INIS)

    Molander, Anders; Ullberg, Mats

    2004-01-01

    Corrosion potential measurements have been performed in Swedish BWRs during 25 years using commercially available monitoring equipment. Today, such measurements are performed on a routine basis in the BWRs on hydrogen water chemistry in Sweden. Measurements are usually performed at several monitoring locations in the plants. During the years, variations in the corrosion potential between different reactor cycles have been observed. Also, the corrosion potential can vary significantly during the reactor year. The changes have not always been easy to explain. Examples of in-plant data are given, demonstrating the need for a better understanding and for improved modeling tools. These examples were used as starting points for developing improved methods for corrosion potential modeling. A new tool recently developed, The Virtual ECP Laboratory, is described and applications to BWR conditions including some unexpected experimental corrosion potential responses are given. (author)

  3. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang

    2014-01-01

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

  4. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  5. Stability analysis of BWR nuclear-coupled thermal-hyraulics using a simple model

    Energy Technology Data Exchange (ETDEWEB)

    Karve, A.A.; Rizwan-uddin; Dorning, J.J. [Univ. of Virginia, Charlottesville, VA (United States)

    1995-09-01

    A simple mathematical model is developed to describe the dynamics of the nuclear-coupled thermal-hydraulics in a boiling water reactor (BWR) core. The model, which incorporates the essential features of neutron kinetics, and single-phase and two-phase thermal-hydraulics, leads to simple dynamical system comprised of a set of nonlinear ordinary differential equations (ODEs). The stability boundary is determined and plotted in the inlet-subcooling-number (enthalpy)/external-reactivity operating parameter plane. The eigenvalues of the Jacobian matrix of the dynamical system also are calculated at various steady-states (fixed points); the results are consistent with those of the direct stability analysis and indicate that a Hopf bifurcation occurs as the stability boundary in the operating parameter plane is crossed. Numerical simulations of the time-dependent, nonlinear ODEs are carried out for selected points in the operating parameter plane to obtain the actual damped and growing oscillations in the neutron number density, the channel inlet flow velocity, and the other phase variables. These indicate that the Hopf bifurcation is subcritical, hence, density wave oscillations with growing amplitude could result from a finite perturbation of the system even where the steady-state is stable. The power-flow map, frequently used by reactor operators during start-up and shut-down operation of a BWR, is mapped to the inlet-subcooling-number/neutron-density (operating-parameter/phase-variable) plane, and then related to the stability boundaries for different fixed inlet velocities corresponding to selected points on the flow-control line. The stability boundaries for different fixed inlet subcooling numbers corresponding to those selected points, are plotted in the neutron-density/inlet-velocity phase variable plane and then the points on the flow-control line are related to their respective stability boundaries in this plane.

  6. 3D modeling of missing pellet surface defects in BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov; Williamson, R.L.; Stafford, D.S.; Novascone, S.R.; Hales, J.D.; Pastore, G.

    2016-10-15

    Highlights: • A global/local analysis procedure for missing pellet surface defects is proposed. • This is applied to defective BWR fuel under blade withdrawal and high power ramp conditions. • Sensitivity of the cladding response to key model parameters is studied. - Abstract: One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can be used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed here. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding

  7. Parameter identification of a BWR nuclear power plant model for use in optimal control

    International Nuclear Information System (INIS)

    Volf, K.

    1976-02-01

    The problem being considered is the modeling of a nuclear power plant for the development of an optimal control system of the plant. Current system identification concepts, combining input/output information with a-priori structural information are employed. Two of the known parameter identification methods i.e., a least squares method and a maximum likelihood technique, are studied as ways of parameter identification from measurement data. A low order state variable stochastic model of a BWR nuclear power plant is presented as an application of this approach. The model consists of a deterministic and a noise part. The deterministic part is formed by simplified modeling of the major plant dynamic phenomena. The moise part models the effects of input random disturbances to the deterministic part and additive measurement noise. Most of the model parameters are assumed to be initially unknown. They are identified using measurement data records. A detailed high order digital computer simulation is used to simulate plant dynamic behaviour since it is not conceivable for experimentation of this kind to be performed on the real nuclear power plant. The identification task consists in adapting the performance of the simple model to the data acquired from this plant simulation ensuring the applicability of the techniques to measurement data acquired directly from the plant. (orig.) [de

  8. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models; SUN-RAH: simulador universitario de nucleoelectrica BWR basado en modelos de orden reducido

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J.B.; Lopez R, A.; Sanchez B, A.; Sanchez S, R.; Hernandez S, A. [DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

    2003-07-01

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  9. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    International Nuclear Information System (INIS)

    Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.; Gauld, Ian C.; Ilas, Germina; Marshall, William BJ J.

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10x10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  10. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models

    International Nuclear Information System (INIS)

    Morales S, J.B.; Lopez R, A.; Sanchez B, A.; Sanchez S, R.; Hernandez S, A.

    2003-01-01

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  11. A nonlinear 3D real-time model for simulation of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.

    1982-02-01

    A nonlinear transient model for BWR nuclear power plants which consists of a 3D-core (subdivided into a number of superboxes, and with parallel flow and subcooled boiling), a top plenum, steam removal and feed water systems and main coolant recirculation pumps is given. The model describes the local core and global plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. The model which forms the basis for the digital code GARLIC-B (Er et al. 82) is aimed to be used on an on-site process computer in parallel to the actual reactor process (or even in predictive mode). Thus, special measures had to be taken into account in order to increase the computational speed and reduce the necessary computer storage. This could be achieved by - separating the neutron and power kinetics from the xenon-iodine dynamics, - treating the neutron kinetics and most of the thermodynamics and hydrodynamics in a pseudostationary way, - developing a special coupling coefficient concept to describe the neutron diffusion, calculating the coupling coefficients from a basic neutron kinetics code, - combining coarse mesh elements into superboxes, taking advantage of the symmetry properties of the core and - applying a sparse matrix technique for solving the resulting algebraic power equation system. (orig.) [de

  12. Investigation on macroscopic cross section model for BWR pin-by-pin core analysis - 118

    International Nuclear Information System (INIS)

    Fujita, T.; Tada, K.; Yamamoto, A.; Yamane, Y.; Kosaka, S.; Hirano, G.

    2010-01-01

    A cross section model used in the pin-by-pin core analysis for BWR is investigated. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of state and history variables that have influences on the cross section and are tabulated prior to the core calculations. Variation of a cross section in a core simulator is classified into two different types, i.e., the instantaneous effect and the history effect. The instantaneous effect is incorporated by the variation of cross section which is caused by the instantaneous change of state variables. For this effect, the exposure, the void fraction, the fuel temperature, the moderator temperature and the control rod are used as indexes. The history effect is the cumulative effect of state variables. We treat this effect with a unified approach using the spectral history. To confirm accuracy of the cross section model, the pin-by-pin fission rate distribution and the k-infinity of fuel assembly which are obtained with the tabulated and the reference cross sections are compared. For the instantaneous effect, the present cross section model well reproduces the reference results for all off-nominal conditions. For the history effect, however, considerable differences both on the pin-by-pin fission rate distribution and the k-infinity are observed at high exposure points. (authors)

  13. Research on PCPV for BWR - physical model as design tool - main results

    International Nuclear Information System (INIS)

    Fumagalli, E.; Verdelli, G.

    1975-01-01

    ISMES (Experimental Institute for Models and Structures) is now carrying out a series of tests on physical models as a part of a research programme sponsored by DSR (Studies and Research Direction) of ENEL (Italian State Electricity Board) on behalf of CPN (Nuclear Design and Construction Centre) of ENEL with the aim to experience a 'Thin'-walled PCPV for 'BWR'. The physical model, together with the mathematical model and the rheological model of the materials, is intended as a meaningful design tool. The mathematical model covers the overall structural design phase, (geometries) and the linear behaviour, whereas the physical model, besides of a global information to be compared with the results of the mathematical model, supplies a number of data as the non-linear behaviour up to failure and local conditions (penetration area etc.) are concerned. The aim of the first phase of this research programme is to make a comparison between the calculation and experiment tests as the thicknesses of the wall and the bottom slab are concerned, whereas the second phase of the research deals with the behaviour of the removable lid and its connection with the main structure. To do this, a model in scale 1:10 has been designed which symmetrically reproduces with respect to the equator, the bottom part of the structure. In the bottom slab the penetrations of the prototype design are reproduced, whereas the upper slab is plain. This paper describes the model, and illustrates the main results, underlining the different behaviour of the upper and bottom slabs up to collapse

  14. A Modeling of BWR-MOX assemblies based on the characteristics method combined with advanced self-shielding models

    International Nuclear Information System (INIS)

    Le Tellier, R.; Hebert, A.; Le Tellier, R.; Santamarina, A.; Litaize, O.

    2008-01-01

    Calculations based on the characteristics method and different self-shielding models are presented for 9 x 9 boiling water reactor (BWR) assemblies fully loaded with mixed-oxide (MOX) fuel. The geometry of these assemblies was recovered from the BASALA experimental program. We have focused our study on three configurations simulating the different voiding conditions that an assembly can undergo in a BWR pressure vessel. A parametric study was carried out with respect to the spatial discretization, the tracking parameters, and the anisotropy order. Comparisons with Monte Carlo calculations in terms of k eff , radiative capture, and fission rates were performed to validate the computational tools. The results are in good agreement between the stochastic and deterministic approaches. The mutual self-shielding model recently introduced within the framework of the Ribon extending self-shielding method appears to be useful for this type of assemblies. Indeed, in the calculation of these MOX benchmarks, the overlapping of resonances, especially between 238 U and 240 Pu, plays an important role due to the spectral strengthening of the flux as the voiding percentage is increased. The method of characteristics is shown to be adequate to perform accurate calculations handling a fine spatial discretization. (authors)

  15. A Non-Linear Digital Computer Model Requiring Short Computation Time for Studies Concerning the Hydrodynamics of the BWR

    Energy Technology Data Exchange (ETDEWEB)

    Reisch, F; Vayssier, G

    1969-05-15

    This non-linear model serves as one of the blocks in a series of codes to study the transient behaviour of BWR or PWR type reactors. This program is intended to be the hydrodynamic part of the BWR core representation or the hydrodynamic part of the PWR heat exchanger secondary side representation. The equations have been prepared for the CSMP digital simulation language. By using the most suitable integration routine available, the ratio of simulation time to real time is about one on an IBM 360/75 digital computer. Use of the slightly different language DSL/40 on an IBM 7044 computer takes about four times longer. The code has been tested against the Eindhoven loop with satisfactory agreement.

  16. Non-Fourier Vernotte-Cattaneo numerical model for heat conduction in a BWR fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Martinez, E.G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [Universidad Autonoma Metropolitana, Area de Ingenieria en Recursos Energeticos, Iztapalapa (Mexico)

    2014-07-01

    A fuel rod mathematical model based on transient heat conduction as constitutive Non-Fourier law for Light Water Reactors (LWRs) transient analysis is presented. The structure of the fuel pellet is affected due to high temperatures and irradiation, which eventually produce fracture or cracks. In principle the fractures are saturated of gas. Then, the Fourier law of the heat conduction is not strictly applicable to describe these phenomena, where the physical properties such as thermal conductivity, heat capacity and density correspond to a heterogeneous material due to gas, and therefore the thermal diffusion process due to molecular transport in the fuel pellet is affected. From the point of view of nuclear reactor safety analysis, the heat transfer from the fuel to the coolant is crucial and superheating of the wall can cause the cladding failure. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The Non-Fourier approach presented in this work eliminates the assumption of an infinite thermal wave speed, therefore time-dependent heat sources were considered in the fuel rod heat transfer model. The numerical experiments in a BWR, show that the Non-Fourier approach is crucial in the pressurization transients such as turbine trip and reactor isolation. (author)

  17. Non-Fourier Vernotte-Cattaneo numerical model for heat conduction in a BWR fuel rod

    International Nuclear Information System (INIS)

    Espinosa-Martinez, E.G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G.

    2014-01-01

    A fuel rod mathematical model based on transient heat conduction as constitutive Non-Fourier law for Light Water Reactors (LWRs) transient analysis is presented. The structure of the fuel pellet is affected due to high temperatures and irradiation, which eventually produce fracture or cracks. In principle the fractures are saturated of gas. Then, the Fourier law of the heat conduction is not strictly applicable to describe these phenomena, where the physical properties such as thermal conductivity, heat capacity and density correspond to a heterogeneous material due to gas, and therefore the thermal diffusion process due to molecular transport in the fuel pellet is affected. From the point of view of nuclear reactor safety analysis, the heat transfer from the fuel to the coolant is crucial and superheating of the wall can cause the cladding failure. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The Non-Fourier approach presented in this work eliminates the assumption of an infinite thermal wave speed, therefore time-dependent heat sources were considered in the fuel rod heat transfer model. The numerical experiments in a BWR, show that the Non-Fourier approach is crucial in the pressurization transients such as turbine trip and reactor isolation. (author)

  18. Results of the Simulator smart against synthetic signals using a model of reduced order of BWR with additive and multiplicative noise; Resultados del simulador smart frente a senales sinteticas utilizando un modelo de orden reducido de BWR con ruido aditivo y multiplicativo

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Montesino, M. E.; Pena, J.; Escriva, A.; Melara, J.

    2011-07-01

    Results of SMART-simulator front of synthetic signals with models of reduced order of BWR with additive and multiplicative noise Under the SMART project, which aims to monitor the signals Cofrentes nuclear plant, we have developed a signal generator of synthetics BWR that will allow together real signals of plant the validation of the monitor.

  19. PVT modeling of reservoir fluids using PC-SAFT EoS and Soave-BWR EoS

    DEFF Research Database (Denmark)

    Yan, Wei; Varzandeh, Farhad; Stenby, Erling Halfdan

    2015-01-01

    non-cubic EoS models, such as the Perturbed Chain Statistical Associating Fluid Theory (PC-SAFT) EoS and the Soave modified Benedict-Webb-Rubin (Soave-BWR) EoS, may partly replace the roles of these classical cubic models in the upstream oil industry. Here, we attempt to make a comparative study...... for the four models. For PVT prediction, the non-cubic models show advantages in some high pressure high temperature (HPHT) fluids but no clear advantages in general, indicating the necessity for further improvement of the characterization procedure....

  20. TRAC-BWR development

    International Nuclear Information System (INIS)

    Weaver, W.L.; Rouhani, S.Z.

    1983-01-01

    The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

  1. Simulation of the BWR experiments CORA-17 and CORA-28 using ATHLET-CD and assessment of BWR modelling. 1{sup st} Technical report. Validation and interpretation of the ATHLET-CD model basis; Simulation der SWR-Versuche CORA-17 und CORA-28 mit dem Programmsystem ATHLET-CD und Bewertung der SWR-Modellbasis. 1. Technischer Fachbericht. Validierung und Interpretation der ATHLET-CD Modellbasis

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, M.; Gremme, F.; Koch, M.K.

    2013-08-15

    . 2.1A. Hydrogen regeneration calculated for CORA-28 agrees well to the measurements, too. In contrast, the hydrogen production in CORA-17 is underestimated during the quenching phase. Possible reasons for these discrepancies are the underestimation of melt oxidation as well as the oxidation of the absorber material B4C since an adequate modeling of these phenomena is not yet implemented in the code. The simulated material relocation of the fuel rods as well as of the BWR components provides plausible results compared to the experiments.

  2. Melts of garnet lherzolite: experiments, models and comparison to melts of pyroxenite and carbonated lherzolite

    Science.gov (United States)

    Grove, Timothy L.; Holbig, Eva S.; Barr, Jay A.; Till, Christy B.; Krawczynski, Michael J.

    2013-01-01

    Phase equilibrium experiments on a compositionally modified olivine leucitite from the Tibetan plateau have been carried out from 2.2 to 2.8 GPa and 1,380–1,480 °C. The experiments-produced liquids multiply saturated with spinel and garnet lherzolite phase assemblages (olivine, orthopyroxene, clinopyroxene and spinel ± garnet) under nominally anhydrous conditions. These SiO2-undersaturated liquids and published experimental data are utilized to develop a predictive model for garnet lherzolite melting of compositionally variable mantle under anhydrous conditions over the pressure range of 1.9–6 GPa. The model estimates the major element compositions of garnet-saturated melts for a range of mantle lherzolite compositions and predicts the conditions of the spinel to garnet lherzolite phase transition for natural peridotite compositions at above-solidus temperatures and pressures. We compare our predicted garnet lherzolite melts to those of pyroxenite and carbonated lherzolite and develop criteria for distinguishing among melts of these different source types. We also use the model in conjunction with a published predictive model for plagioclase and spinel lherzolite to characterize the differences in major element composition for melts in the plagioclase, spinel and garnet facies and develop tests to distinguish between melts of these three lherzolite facies based on major elements. The model is applied to understand the source materials and conditions of melting for high-K lavas erupted in the Tibetan plateau, basanite–nephelinite lavas erupted early in the evolution of Kilauea volcano, Hawaii, as well as younger tholeiitic to alkali lavas from Kilauea.

  3. Application of Continuous and Structural ARMA modeling for noise analysis of a BWR coupled core and plant instability event

    International Nuclear Information System (INIS)

    Demeshko, M.; Dokhane, A.; Washio, T.; Ferroukhi, H.; Kawahara, Y.; Aguirre, C.

    2015-01-01

    Highlights: • We demonstrate the first application of a novel CSARMA method. • We analyze the instability occurred in a Swiss BWR plant during power ascension. • Benchmarked the results against STP analysis. • The CSARMA results are consistent with the background physics and the STP results. • The instability was caused by disturbances in the pressure control system. - Abstract: This paper presents a first application of a novel Continuous and Structural Autoregressive Moving Average (CSARMA) modeling approach to BWR noise analysis. The CSARMA approach derives a unique representation of the system dynamics by more robust and reliable canonical models as basis for signal analysis in general and for reactor diagnostics in particular. In this paper, a stability event that occurred in a Swiss BWR plant during power ascension phase is analyzed as well as the time periods that preceded and followed the event. Focusing only on qualitative trends at this stage, the obtained results clearly indicate a different dynamical state during the unstable event compared to the two other stable periods. Also, they could be interpreted as pointing out a disturbance in the pressure control system as primary cause for the event. To benchmark these findings, the frequency-domain based signal transmission-path (STP) method is also applied. And with the STP method, we obtained similar relationships as mentioned above. This consistency between both methods can be considered as being a confirmation that the event was caused by a pressure control system disturbance and not induced by the core. Also, it is worth noting that the STP analysis failed to catch the relations among the processes during the stable periods, that were clearly indicated by the CSARMA method, since the last uses more precise models as basis

  4. A multi-component evaporation model for beam melting processes

    Science.gov (United States)

    Klassen, Alexander; Forster, Vera E.; Körner, Carolin

    2017-02-01

    In additive manufacturing using laser or electron beam melting technologies, evaporation losses and changes in chemical composition are known issues when processing alloys with volatile elements. In this paper, a recently described numerical model based on a two-dimensional free surface lattice Boltzmann method is further developed to incorporate the effects of multi-component evaporation. The model takes into account the local melt pool composition during heating and fusion of metal powder. For validation, the titanium alloy Ti-6Al-4V is melted by selective electron beam melting and analysed using mass loss measurements and high-resolution microprobe imaging. Numerically determined evaporation losses and spatial distributions of aluminium compare well with experimental data. Predictions of the melt pool formation in bulk samples provide insight into the competition between the loss of volatile alloying elements from the irradiated surface and their advective redistribution within the molten region.

  5. Nuclear-coupled thermal-hydraulic nonlinear stability analysis using a novel BWR reduced order model. Pt. 1. The effects of using drift flux versus homogeneous equilibrium models

    International Nuclear Information System (INIS)

    Dokhane, A.; Henning, D.; Chawla, R.; Rizwan-Uddin

    2003-01-01

    BWR stability analysis at PSI, as at other research centres, is usually carried out employing complex system codes. However, these do not allow a detailed investigation of the complete manifold of all possible solutions of the associated nonlinear differential equation set. A novel analytical, reduced order model for BWR stability has been developed at PSI, in several successive steps. In the first step, the thermal-hydraulic model was used for studying the thermal-hydraulic instabilities. A study was then conducted of the one-channel nuclear-coupled thermal-hydraulic dynamics in a BWR by adding a simple point kinetic model for neutron kinetics and a model for the fuel heat conduction dynamics. In this paper, a two-channel nuclear-coupled thermal-hydraulic model is introduced to simulate the out-of phase oscillations in a BWR. This model comprises three parts: spatial mode neutron kinetics with the fundamental and fist azimuthal modes; fuel heat conduction dynamics; and thermal-hydraulics model. This present model is an extension of the Karve et al. model i.e., a drift flux model is used instead of the homogeneous equilibrium model for two-phase flow, and lambda modes are used instead of the omega modes for the neutron kinetics. This two-channel model is employed in stability and bifurcation analyses, carried out using the bifurcation code BIFDD. The stability boundary (SB) and the nature of the Poincare-Andronov-Hopf bifurcation (PAF-B) are determined and visualized in a suitable two-dimensional parameter/state space. A comparative study of the homogeneous equilibrium model (HEM) and the drift flux model (DFM) is carried out to investigate the effects of the DFM parameters the void distribution parameter C 0 and the drift velocity V gi -on the SB, the nature of PAH bifurcation, and on the type of oscillation mode (in-phase or out-of-phase). (author)

  6. A 2D double-porosity model for melting and melt migration beneath mid-oceanic ridges

    Science.gov (United States)

    Liu, B.; Liang, Y.; Parmentier, E.

    2017-12-01

    Several lines of evidence suggest that the melting and melt extraction region of the MORB mantle is heterogeneous consisting of an interconnected network of high permeability dunite channels in a low porosity harzburgite or lherzolite matrix. In principle, one can include channel formation into the tectonic-scale geodynamic models by solving conservation equations for a chemically reactive and viscously deformable porous medium. Such an approach eventually runs into computational limitations such as resolving fractal-like channels that have a spectrum of width. To better understand first order features of melting and melt-rock interaction beneath MOR, we have formulated a 2D double porosity model in which we treat the triangular melting region as two overlapping continua occupied by the low-porosity matrix and interconnected high-porosity channels. We use melt productivity derived from a thermodynamic model and melt suction rate to close our problem. We use a high-order accurate numerical method to solve the conservation equations in 2D for porosity, solid and melt velocities and concentrations of chemical tracers in the melting region. We carry out numerical simulations to systematically study effects of matrix-to-channel melt suction and spatially distributed channels on the distributions of porosity and trace element and isotopic ratios in the melting region. For near fractional melting with 10 vol% channel in the melting region, the flow field of the matrix melt follows closely to that of the solid because the small porosity (exchange between the melt and the solid. The smearing effect can be approximated by dispersion coefficient. For slowly diffusing trace elements (e.g., LREE and HFSE), the melt migration induced dispersion can be as effective as thermal diffusion. Therefore, sub-kilometer scale heterogeneities of Nd and Hf isotopes are significantly damped or homogenized in the melting region.

  7. Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

    International Nuclear Information System (INIS)

    Adamsson, Carl; Le Corre, Jean-Marie

    2011-01-01

    Highlights: → The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. → A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. → MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. → The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. → The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the

  8. Viscosity of Heterogeneous Silicate Melts: A Non-Newtonian Model

    Science.gov (United States)

    Liu, Zhuangzhuang; Blanpain, Bart; Guo, Muxing

    2017-12-01

    The recently published viscosity data of heterogeneous silicate melts with well-documented structure and experimental conditions are critically re-analyzed and tabulated. By using these data, a non-Newtonian viscosity model incorporating solid fraction, solid shape, and shear rate is proposed on the basis of the power-law equation. This model allows calculating the viscosity of the heterogeneous silicate melts with solid fraction up to 34 vol pct. The error between the calculated and measured data is evaluated to be 32 pct, which is acceptable considering the large error in viscosity measurement of the completely liquid silicate melt.

  9. Exxon Nuclear WREM-based NJP-BWR ECCS evaluation model and example application to the Oyster Creek Plant

    International Nuclear Information System (INIS)

    Krysinski, T.L.; Bjornard, T.A.; Steves, L.H.

    1975-01-01

    A proposed integrated ECCS model for non-jet pump boiling water reactors is presented, using the RELAP4-EM/BLOWDOWN and RELAP4-EM/SMALL BREAK portions of the Exxon Nuclear WREM-based Generic PWR Evaluation Model coupled with the ENC NJP-BWR Fuel Heatup Model. The results of the application of the proposed model to Oyster Creek are summarized. The results of the break size sensitivity study using the proposed model for the Oyster Creek Plant are presented. The application of the above results yielded the MAPLHGR curves. Included are a description of the proposed non-jet pump boiling water reaction evaluation model, justification of its conformance with TOCFR50, Appendix K, the adopted Oyster Creek plant model, and results of the analysis and sensitivity studies. (auth)

  10. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH

    International Nuclear Information System (INIS)

    Barron A, I.

    2005-01-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  11. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor

    International Nuclear Information System (INIS)

    Victoria R, M.A.; Morales S, J.B.

    2005-01-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  12. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

  13. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report

    International Nuclear Information System (INIS)

    Tentner, A.

    2009-01-01

    A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

  14. Diffusion of hydrous species in model basaltic melt

    Science.gov (United States)

    Zhang, Li; Guo, Xuan; Wang, Qinxia; Ding, Jiale; Ni, Huaiwei

    2017-10-01

    Water diffusion in Fe-free model basaltic melt with up to 2 wt% H2O was investigated at 1658-1846 K and 1 GPa in piston-cylinder apparatus using both hydration and diffusion couple techniques. Diffusion profiles measured by FTIR are consistent with a model in which both molecular H2O (H2Om) and hydroxyl (OH) contribute to water diffusion. OH diffusivity is roughly 13% of H2Om diffusivity, showing little dependence on temperature or water concentration. Water diffusion is dominated by the motion of OH until total H2O (H2Ot) concentration reaches 1 wt%. The dependence of apparent H2Ot diffusivity on H2Ot concentration appears to be overestimated by a previous study on MORB melt, but H2Ot diffusivity at 1 wt% H2Ot in basaltic melt is still greater than those in rhyolitic to andesitic melts. The appreciable contribution of OH to water diffusion in basaltic melt can be explained by enhanced mobility of OH, probably associated with the development of free hydroxyl bonded with network-modifying cations, as well as higher OH concentration. Calculation based on the Nernst-Einstein equation demonstrates that OH may serve as an effective charge carrier in hydrous basaltic melt, which could partly account for the previously observed strong influence of water on electrical conductivity of basaltic melt.

  15. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  16. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  17. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

    2005-07-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  18. Modeling of nuclear waste disposal by rock melting

    International Nuclear Information System (INIS)

    Heuze, F.E.

    1982-04-01

    Today, the favored option for disposal of high-level nuclear wastes is their burial in mined caverns. As an alternative, the concept of deep disposal by rock melting (DRM) also has received some attention. DRM entails the injection of waste, in a cavity or borehole, 2 to 3 kilometers down in the earth crust. Granitic rocks are the prime candidate medium. The high thermal loading initially will melt the rock surrounding the waste. Following resolidification, a rock/waste matrix is formed, which should provide isolation for many years. The complex thermal, mechanical, and hydraulic aspects of DRM can be studied best by means of numerical models. The models must accommodate the coupling of the physical processes involved, and the temperature dependency of the granite properties, some of which are subject to abrupt discontinuities, during α-β phase transition and melting. This paper outlines a strategy for such complex modeling

  19. Modeling the kinetics of volatilization from glass melts

    NARCIS (Netherlands)

    Beerkens, R.G.C.

    2001-01-01

    A model description for the evaporation kinetics from glass melts in direct contact with static atmospheres or flowing gas phases is presented. The derived models and equations are based on the solution of the second Ficks' diffusion law and quasi-steady-state mass transfer relations, taking into

  20. Simplification of neural network model for predicting local power distributions of BWR fuel bundle using learning algorithm with forgetting

    International Nuclear Information System (INIS)

    Tanabe, Akira; Yamamoto, Toru; Shinfuku, Kimihiro; Nakamae, Takuji; Nishide, Fusayo.

    1995-01-01

    Previously a two-layered neural network model was developed to predict the relation between fissile enrichment of each fuel rod and local power distribution in a BWR fuel bundle. This model was obtained intuitively based on 33 patterns of training signals after an intensive survey of the models. Recently, a learning algorithm with forgetting was reported to simplify neural network models. It is an interesting subject what kind of model will be obtained if this algorithm is applied to the complex three-layered model which learns the same training signals. A three-layered model which is expanded to have direct connections between the 1st and the 3rd layer elements has been constructed and the learning method of normal back propagation was applied first to this model. The forgetting algorithm was then added to this learning process. The connections concerned with the 2nd layer elements disappeared and the 2nd layer has become unnecessary. It took a longer computing time by an order to learn the same training signals than the simple back propagation, but the two-layered model was obtained autonomously from the expanded three-layered model. (author)

  1. Modeling of the water gap in BWR fuel elements using SCALE/TRITON; Modellierung des Wasserspalts bei SWR-BE mit SCALE/TRITON

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S.; Chernykh, M. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2012-11-01

    The authors show that an adequate modeling of the water gap in BWR fuel element models using the code TRITON requires an explicit consideration of the Dancoff factors. The analysis of three modeling options reveals that considering the moderating effects of the water gap coolant for the peripheral fuel elements the resulting deviations of the U-235 and Pu-239 concentrations are significantly reduced. The increased temporal calculation efforts are justified with respect to the burnup credits for criticality safety analyses.

  2. Bubble removal and sand dissolution in an electrically heated glass melting channel with defined melt flow examined by mathematical modelling

    Czech Academy of Sciences Publication Activity Database

    Hrbek, L.; Kocourková, P.; Jebavá, Marcela; Cincibusová, P.; Němec, Lubomír

    2017-01-01

    Roč. 456, JAN 15 (2017), s. 101-113 ISSN 0022-3093 Institutional support: RVO:67985891 Keywords : glass melt flow * mathematical modelling * energy distribution * space utilization * melting performance Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass OBOR OECD: Ceramics Impact factor: 2.124, year: 2016

  3. Numerical Model based Reliability Estimation of Selective Laser Melting Process

    DEFF Research Database (Denmark)

    Mohanty, Sankhya; Hattel, Jesper Henri

    2014-01-01

    Selective laser melting is developing into a standard manufacturing technology with applications in various sectors. However, the process is still far from being at par with conventional processes such as welding and casting, the primary reason of which is the unreliability of the process. While...... of the selective laser melting process. A validated 3D finite-volume alternating-direction-implicit numerical technique is used to model the selective laser melting process, and is calibrated against results from single track formation experiments. Correlation coefficients are determined for process input...... parameters such as laser power, speed, beam profile, etc. Subsequently, uncertainties in the processing parameters are utilized to predict a range for the various outputs, using a Monte Carlo method based uncertainty analysis methodology, and the reliability of the process is established....

  4. Lattice Boltzmann model for melting with natural convection

    International Nuclear Information System (INIS)

    Huber, Christian; Parmigiani, Andrea; Chopard, Bastien; Manga, Michael; Bachmann, Olivier

    2008-01-01

    We develop a lattice Boltzmann method to couple thermal convection and pure-substance melting. The transition from conduction-dominated heat transfer to fully-developed convection is analyzed and scaling laws and previous numerical results are reproduced by our numerical method. We also investigate the limit in which thermal inertia (high Stefan number) cannot be neglected. We use our results to extend the scaling relations obtained at low Stefan number and establish the correlation between the melting front propagation and the Stefan number for fully-developed convection. We conclude by showing that the model presented here is particularly well-suited to study convection melting in geometrically complex media with many applications in geosciences

  5. Residence time modeling of hot melt extrusion processes.

    Science.gov (United States)

    Reitz, Elena; Podhaisky, Helmut; Ely, David; Thommes, Markus

    2013-11-01

    The hot melt extrusion process is a widespread technique to mix viscous melts. The residence time of material in the process frequently determines the product properties. An experimental setup and a corresponding mathematical model were developed to evaluate residence time and residence time distribution in twin screw extrusion processes. The extrusion process was modeled as the convolution of a mass transport process described by a Gaussian probability function, and a mixing process represented by an exponential function. The residence time of the extrusion process was determined by introducing a tracer at the extruder inlet and measuring the tracer concentration at the die. These concentrations were fitted to the residence time model, and an adequate correlation was found. Different parameters were derived to characterize the extrusion process including the dead time, the apparent mixing volume, and a transport related axial mixing. A 2(3) design of experiments was performed to evaluate the effect of powder feed rate, screw speed, and melt viscosity of the material on the residence time. All three parameters affect the residence time of material in the extruder. In conclusion, a residence time model was developed to interpret experimental data and to get insights into the hot melt extrusion process. Copyright © 2013 Elsevier B.V. All rights reserved.

  6. Population balance modelling of fluidized bed melt granulation: an overview

    NARCIS (Netherlands)

    Tan, H.S.; Goldschmidt, M.J.V.; Boerefijn, R.; Hounslow, M.J.; Salman, A.; Kuipers, J.A.M.

    2005-01-01

    This paper presents an overview of the work undertaken by our group to identify and quantify the rates processes active in fluidized bed melt granulation (FBMG). The process involves the identification and development of physically representative models to mechanistically describe FBMG using both

  7. Computationally efficient thermal-mechanical modelling of selective laser melting

    NARCIS (Netherlands)

    Yang, Y.; Ayas, C.; Brabazon, Dermot; Naher, Sumsun; Ul Ahad, Inam

    2017-01-01

    The Selective laser melting (SLM) is a powder based additive manufacturing (AM) method to produce high density metal parts with complex topology. However, part distortions and accompanying residual stresses deteriorates the mechanical reliability of SLM products. Modelling of the SLM process is

  8. A Structural Molar Volume Model for Oxide Melts Part III: Fe Oxide-Containing Melts

    Science.gov (United States)

    Thibodeau, Eric; Gheribi, Aimen E.; Jung, In-Ho

    2016-04-01

    As part III of this series, the model is extended to iron oxide-containing melts. All available experimental data in the FeO-Fe2O3-Na2O-K2O-MgO-CaO-MnO-Al2O3-SiO2 system were critically evaluated based on the experimental condition. The variations of FeO and Fe2O3 in the melts were taken into account by using FactSage to calculate the Fe2+/Fe3+ distribution. The molar volume model with unary and binary model parameters can be used to predict the molar volume of the molten oxide of the Li2O-Na2O-K2O-MgO-CaO-MnO-PbO-FeO-Fe2O3-Al2O3-SiO2 system in the entire range of compositions, temperatures, and oxygen partial pressures from Fe saturation to 1 atm pressure.

  9. Evaporation experiments and modelling for glass melts

    NARCIS (Netherlands)

    Limpt, J.A.C. van; Beerkens, R.G.C.

    2007-01-01

    A laboratory test facility has been developed to measure evaporation rates of different volatile components from commercial and model glass compositions. In the set-up the furnace atmosphere, temperature level, gas velocity and batch composition are controlled. Evaporation rates have been measured

  10. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  11. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    Science.gov (United States)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  12. Enthalpy model for heating, melting, and vaporization in laser ablation

    OpenAIRE

    Vasilios Alexiades; David Autrique

    2010-01-01

    Laser ablation is used in a growing number of applications in various areas including medicine, archaeology, chemistry, environmental and materials sciences. In this work the heat transfer and phase change phenomena during nanosecond laser ablation of a copper (Cu) target in a helium (He) background gas at atmospheric pressure are presented. An enthalpy model is outlined, which accounts for heating, melting, and vaporization of the target. As far as we know, this is the first model th...

  13. Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

    International Nuclear Information System (INIS)

    Griffin, F.P.

    1995-01-01

    A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B 4 C, stainless steel, and Zircaloy. This paper describes improvements that have been made to the BWR control blade/channel box model during 1994 and 1995. These improvements include new capabilities that represent the relocation of molten material in a more realistic manner and modifications that improve the usability of the code by reducing the frequency of code failures. This paper also describes a SCDAP/RELAP5 assessment calculation for the Browns Ferry Nuclear Plant design based upon a short-term station blackout accident sequence

  14. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  15. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  16. An empirical model for the melt viscosity of polymer blends

    International Nuclear Information System (INIS)

    Dobrescu, V.

    1981-01-01

    On the basis of experimental data for blends of polyethylene with different polymers an empirical equation is proposed to describe the dependence of melt viscosity of blends on component viscosities and composition. The model ensures the continuity of viscosity vs. composition curves throughout the whole composition range, the possibility of obtaining extremum values higher or lower than the viscosities of components, allows the calculation of flow curves of blends from the flow curves of components and their volume fractions. (orig.)

  17. Measurement and modeling of two-phase flow parameters in scaled 8 Multiplication-Sign 8 BWR rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Yang, X.; Schlegel, J.P.; Liu, Y.; Paranjape, S.; Hibiki, T. [School of Nuclear Engineering, Purdue University, 400 Central Dr., West Lafayette, IN 47907-2017 (United States); Ishii, M., E-mail: ishii@purdue.edu [School of Nuclear Engineering, Purdue University, 400 Central Dr., West Lafayette, IN 47907-2017 (United States)

    2012-04-15

    Highlights: Black-Right-Pointing-Pointer Grid spacers have a significant but not well understood effect on flow behavior and development. Black-Right-Pointing-Pointer Two different length scales are present in rod bundles, which must be accounted for in modeling. Black-Right-Pointing-Pointer An easy-to-implement empirical model has been developed for the two-phase friction multiplier. - Abstract: The behavior of reactor systems is predicted using advanced computational codes in order to determine the safety characteristics of the system during various accidents and to determine the performance characteristics of the reactor. These codes generally utilize the two-fluid model for predictions of two-phase flows, as this model is the most accurate and detailed model which is currently practical for predicting large-scale systems. One of the weaknesses of this approach however is the need to develop constitutive models for various quantities. Of specific interest are the models used in the prediction of void fraction and pressure drop across the rod bundle due to their importance in new Natural Circulation Boiling Water Reactor (NCBWR) designs, where these quantities determine the coolant flow rate through the core. To verify the performance of these models and expand the existing experimental database, data has been collected in an 8 Multiplication-Sign 8 rod bundle which is carefully scaled from actual BWR geometry and includes grid spacers to maintain rod spacing. While these spacer grids are 'generic', their inclusion does provide valuable data for analysis of the effect of grid spacers on the flow. In addition to pressure drop measurements the area-averaged void fraction has been measured by impedance void meters and local conductivity probes have been used to measure the local void fraction and interfacial area concentration in the bundle subchannels. Experimental conditions covered a wide range of flow rates and void fractions up to 80%.

  18. Melt coolability modeling and comparison to MACE test results

    International Nuclear Information System (INIS)

    Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.

    1992-01-01

    An important question in the assessment of severe accidents in light water nuclear reactors is the ability of water to quench a molten corium-concrete interaction and thereby terminate the accident progression. As part of the Melt Attack and Coolability Experiment (MACE) Program, phenomenological models of the corium quenching process are under development. The modeling approach considers both bulk cooldown and crust-limited heat transfer regimes, as well as criteria for the pool thermal hydraulic conditions which separate the two regimes. The model is then compared with results of the MACE experiments

  19. Numerical simulation of hot-melt extrusion processes for amorphous solid dispersions using model-based melt viscosity.

    Science.gov (United States)

    Bochmann, Esther S; Steffens, Kristina E; Gryczke, Andreas; Wagner, Karl G

    2018-03-01

    Simulation of HME processes is a valuable tool for increased process understanding and ease of scale-up. However, the experimental determination of all required input parameters is tedious, namely the melt rheology of the amorphous solid dispersion (ASD) in question. Hence, a procedure to simplify the application of hot-melt extrusion (HME) simulation for forming amorphous solid dispersions (ASD) is presented. The commercial 1D simulation software Ludovic ® was used to conduct (i) simulations using a full experimental data set of all input variables including melt rheology and (ii) simulations using model-based melt viscosity data based on the ASDs glass transition and the physical properties of polymeric matrix only. Both types of HME computation were further compared to experimental HME results. Variation in physical properties (e.g. heat capacity, density) and several process characteristics of HME (residence time distribution, energy consumption) among the simulations and experiments were evaluated. The model-based melt viscosity was calculated by using the glass transition temperature (T g ) of the investigated blend and the melt viscosity of the polymeric matrix by means of a T g -viscosity correlation. The results of measured melt viscosity and model-based melt viscosity were similar with only few exceptions, leading to similar HME simulation outcomes. At the end, the experimental effort prior to HME simulation could be minimized and the procedure enables a good starting point for rational development of ASDs by means of HME. As model excipients, Vinylpyrrolidone-vinyl acetate copolymer (COP) in combination with various APIs (carbamazepine, dipyridamole, indomethacin, and ibuprofen) or polyethylene glycol (PEG 1500) as plasticizer were used to form the ASDs. Copyright © 2017 Elsevier B.V. All rights reserved.

  20. Model of fracture of metal melts and the strength of melts under dynamic conditions

    International Nuclear Information System (INIS)

    Mayer, P. N.; Mayer, A. E.

    2015-01-01

    The development of a continuum model of deformation and fracture of melts is needed for the description of the behavior of metals in extreme states, in particular, under high-current electron and ultrashort laser irradiation. The model proposed includes the equations of mechanics of a two-phase continuum and the equations of the kinetics of phase transitions. The change (exchange) of the volumes of dispersed and carrier phases and of the number of dispersed particles is described, and the energy and mass exchange between the phases due to phase transitions is taken into account. Molecular dynamic (MD) calculations are carried out with the use of the LAMMPS program. The continuum model is verified by MD, computational, and experimental data. The strength of aluminum, copper, and nickel is determined at various temperatures and strain rates. It is shown that an increase in the strain rate leads to an increase in the strength of a liquid metal, while an increase in temperature leads to a decrease in its strength

  1. Model of fracture of metal melts and the strength of melts under dynamic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, P. N., E-mail: polina.nik@mail.ru; Mayer, A. E., E-mail: mayer@csu.ru [Chelyabinsk State University (Russian Federation)

    2015-07-15

    The development of a continuum model of deformation and fracture of melts is needed for the description of the behavior of metals in extreme states, in particular, under high-current electron and ultrashort laser irradiation. The model proposed includes the equations of mechanics of a two-phase continuum and the equations of the kinetics of phase transitions. The change (exchange) of the volumes of dispersed and carrier phases and of the number of dispersed particles is described, and the energy and mass exchange between the phases due to phase transitions is taken into account. Molecular dynamic (MD) calculations are carried out with the use of the LAMMPS program. The continuum model is verified by MD, computational, and experimental data. The strength of aluminum, copper, and nickel is determined at various temperatures and strain rates. It is shown that an increase in the strain rate leads to an increase in the strength of a liquid metal, while an increase in temperature leads to a decrease in its strength.

  2. Enthalpy model for heating, melting, and vaporization in laser ablation

    Directory of Open Access Journals (Sweden)

    Vasilios Alexiades

    2010-09-01

    Full Text Available Laser ablation is used in a growing number of applications in various areas including medicine, archaeology, chemistry, environmental and materials sciences. In this work the heat transfer and phase change phenomena during nanosecond laser ablation of a copper (Cu target in a helium (He background gas at atmospheric pressure are presented. An enthalpy model is outlined, which accounts for heating, melting, and vaporization of the target. As far as we know, this is the first model that connects the thermodynamics and underlying kinetics of this challenging phase change problem in a self-consistent way.

  3. Gas Atomization of Aluminium Melts: Comparison of Analytical Models

    Directory of Open Access Journals (Sweden)

    Georgios Antipas

    2012-06-01

    Full Text Available A number of analytical models predicting the size distribution of particles during atomization of Al-based alloys by N2, He and Ar gases were compared. Simulations of liquid break up in a close coupled atomizer revealed that the finer particles are located near the center of the spray cone. Increasing gas injection pressures led to an overall reduction of particle diameters and caused a migration of the larger powder particles towards the outer boundary of the flow. At sufficiently high gas pressures the spray became monodisperse. The models also indicated that there is a minimum achievable mean diameter for any melt/gas system.

  4. Heat transfer modelling and stability analysis of selective laser melting

    International Nuclear Information System (INIS)

    Gusarov, A.V.; Yadroitsev, I.; Bertrand, Ph.; Smurov, I.

    2007-01-01

    The process of direct manufacturing by selective laser melting basically consists of laser beam scanning over a thin powder layer deposited on a dense substrate. Complete remelting of the powder in the scanned zone and its good adhesion to the substrate ensure obtaining functional parts with improved mechanical properties. Experiments with single-line scanning indicate, that an interval of scanning velocities exists where the remelted tracks are uniform. The tracks become broken if the scanning velocity is outside this interval. This is extremely undesirable and referred to as the 'balling' effect. A numerical model of coupled radiation and heat transfer is proposed to analyse the observed instability. The 'balling' effect at high scanning velocities (above ∼20 cm/s for the present conditions) can be explained by the Plateau-Rayleigh capillary instability of the melt pool. Two factors stabilize the process with decreasing the scanning velocity: reducing the length-to-width ratio of the melt pool and increasing the width of its contact with the substrate

  5. Application of multicomponent medium model for numerical simulation of reactor element melting and melt relocation under severe accidents

    International Nuclear Information System (INIS)

    Vladimir Ya Kumaev

    2005-01-01

    Full text of publication follows: Numerical simulation of the melting processes is necessary in substantiating the safety of new generation reactors to determine the quantitative characteristics of the melt formed, destruction of reactor vessel and components, melt interaction processes in the melt localization systems (MLS), formation and transport of hydrogen, radioactive aerosols under severe accidents. The results of computations will be applied in developing the procedures for severe accident management and mitigation of its consequences and designing melt localization systems. The report is devoted to the development and application of the two-dimensional and three-dimensional versions of the DINCOR code intended for numerical simulation of the thermal hydraulic processes in a multicomponent medium with solid-liquid phase changes. The basic set of equations of multicomponent medium is presented. The numerical method to solve the governing equations is discussed. Some examples of two-dimensional code applications are presented. The experience of application of the code has shown that joint calculations of hydrodynamics, heat transfer, stratification and chemical interaction enable the process description accuracy to be significantly increased and the number of initial experimental data to be reduced. The multicomponent medium model can be used as the base for the development of a three-dimensional version of the code. At the same time, it was established that the models being used need be further developed. The most important problems are the following: -development of the local mathematical models of liquefaction and solidification of materials under front melting and melting due to the action of internal sources; -development of the model of incompressible components separation; -development of the models of dissolution and chemical interaction of multicomponent medium components. In conclusion possible verification of the computer code is discussed. (author)

  6. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

  7. Discrete elastic model for two-dimensional melting.

    Science.gov (United States)

    Lansac, Yves; Glaser, Matthew A; Clark, Noel A

    2006-04-01

    We present a network model for the study of melting and liquid structure in two dimensions, the first in which the presence and energy of topological defects (dislocations and disclinations) and of geometrical defects (elemental voids) can be independently controlled. Interparticle interaction is via harmonic springs and control is achieved by Monte Carlo moves which springs can either be orientationally "flipped" between particles to generate topological defects, or can be "popped" in force-free shape, to generate geometrical defects. With the geometrical defects suppressed the transition to the liquid phase occurs via disclination unbinding, as described by the Kosterlitz-Thouless-Halperin-Nelson-Young model and found in soft potential two-dimensional (2D) systems, such as the dipole-dipole potential [H. H. von Grünberg, Phys. Rev. Lett. 93, 255703 (2004)]. By contrast, with topological defects suppressed, a disordering transition, the Glaser-Clark condensation of geometrical defects [M. A. Glaser and N. A. Clark, Adv. Chem. Phys. 83, 543 (1993); M. A. Glaser, (Springer-Verlag, Berlin, 1990), Vol. 52, p. 141], produces a state that accurately characterizes the local liquid structure and first-order melting observed in hard-potential 2D systems, such as hard disk and the Weeks-Chandler-Andersen (WCA) potentials (M. A. Glaser and co-workers, see above). Thus both the geometrical and topological defect systems play a role in melting. The present work introduces a system in which the relative roles of topological and geometrical defects and their interactions can be explored. We perform Monte Carlo simulations of this model in the isobaric-isothermal ensemble, and present the phase diagram as well as various thermodynamic, statistical, and structural quantities as a function of the relative populations of geometrical and topological defects. The model exhibits a rich phase behavior including hexagonal and square crystals, expanded crystal, dodecagonal quasicrystal

  8. Modeling of evaporation processes in glass melting furnaces

    NARCIS (Netherlands)

    Limpt, van J.A.C.

    2007-01-01

    The majority of glass furnaces worldwide, apply fossil fuel combustion to transfer heat directly by radiation from the combustion processes to the melting batch and glass melt. During these high temperature melting processes, some glass components, such as: sodium, potassium, boron and lead species

  9. Modeling and simulation of Si crystal growth from melt

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Lijun; Liu, Xin; Li, Zaoyang [National Engineering Research Center for Fluid Machinery and Compressors, School of Energy and Power Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi 710049 (China); Miyazawa, Hiroaki; Nakano, Satoshi; Kakimoto, Koichi [Research Institute for Applied Mechanics, Kyushu University, Kasuga 816-8580 (Japan)

    2009-07-01

    A numerical simulator was developed with a global model of heat transfer for any crystal growth taking place at high temperature. Convective, conductive and radiative heat transfers in the furnace are solved together in a conjugated way by a finite volume method. A three-dimensional (3D) global model was especially developed for simulation of heat transfer in any crystal growth with 3D features. The model enables 3D global simulation be conducted with moderate requirement of computer resources. The application of this numerical simulator to a CZ growth and a directional solidification process for Si crystals, the two major production methods for crystalline Si for solar cells, was introduced. Some typical results were presented, showing the importance and effectiveness of numerical simulation in analyzing and improving these kinds of Si crystal growth processes from melt. (copyright 2009 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  10. Volume dependence of the melting temperature for alkali metals with Debye's model

    International Nuclear Information System (INIS)

    Soma, T.; Kagaya, H.M.; Nishigaki, M.

    1983-01-01

    Using the volume dependence of the Grueneisen constant at higher temperatures, the volume effect on the melting temperature of alkali metals is studied by Lindeman's melting law and Debye's model. The obtained melting curve increases as a function of the compressed volume and shows the maximum of the melting point at the characteristic volume. The resultant data are qualitatively in agreement with the observed tendency for alkali metals. (author)

  11. 2D model for melt progression through rods and debris

    International Nuclear Information System (INIS)

    Fichot, F.

    2001-01-01

    During the degradation of a nuclear core in a severe accident scenario, the high temperatures reached lead to the melting of materials. The formation of liquid mixtures at various elevations is followed by the flow of molten materials through the core. Liquid mixture may flow under several configurations: axial relocation along the rods, horizontal motion over a plane surface such as the core support plate or a blockage of material, 2D relocation through a debris bed, etc.. The two-dimensional relocation of molten material through a porous debris bed, implemented for the simulation of late degradation phases, has opened a new way to the elaboration of the relocation model for the flow of liquid mixture along the rods. It is based on a volume averaging method, where wall friction and capillary effects are taken into account by introducing effective coefficients to characterize the solid matrix (rods, grids, debris, etc.). A local description of the liquid flow is necessary to derive the effective coefficients. Heat transfers are modelled in a similar way. The derivation of the conservation equations for the liquid mixture falling flow (momentum) in two directions (axial and radial-horizontal) and for the heat exchanges (energy) are the main points of this new model for simulating melt progression. In this presentation, the full model for the relocation and solidification of liquid materials through a rod bundle or a debris bed is described. It is implemented in the ICARE/CATHARE code, developed by IPSN in Cadarache. The main improvements and advantages of the new model are: A single formulation for liquid mixture relocation, in 2D, either through a rod bundle or a porous debris bed, Extensions to complex structures (grids, by-pass, etc..), The modeling of relocation of a liquid mixture over plane surfaces. (author)

  12. MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

    International Nuclear Information System (INIS)

    Tautges, T.J.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data

  13. Interpretation of the results of the CORA-33 dry core BWR test

    International Nuclear Information System (INIS)

    Ott, L.J.; Hagen, S.

    1993-01-01

    All BWR degraded core experiments performed prior to CORA-33 were conducted under ''wet'' core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ''dry'' core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ''dry'' core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a ''dry'' BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ''dry'' core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed

  14. Investigations of model polymers: Dynamics of melts and statics of a long chain in a dilute melt of shorter chains

    International Nuclear Information System (INIS)

    Bishop, M.; Ceperley, D.; Frisch, H.L.; Kalos, M.H.

    1982-01-01

    We report additional results on a simple model of polymers, namely the diffusion in concentrated polymer systems and the static properties of one long chain in a dilute melt of shorter chains. It is found, for the polymer sizes and time scales amenable to our computer calculations, that there is as yet no evidence for a ''reptation'' regime in a melt. There is some indication of reptation in the case of a single chain moving through fixed obstacles. No statistically significant effect of the change, from excluded volume behavior of the long chain to ideal behavior as the shorter chains grow, is observed

  15. Development of a dynamic model of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Nonboel, E.

    1975-12-01

    A description is given of a one-dimensional steady-state model of a high-pressure steam turbine, a low-pressure steam turbine, a moisture separator, a reheater, a condenser, feedwater heaters and feedwater pump for a nuclear power plant. The model is contained in the program ''TURBPLANT''. The dynamic part of this model is presented in part II of this report. (author)

  16. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

  17. BWR MARK I pressure suppression pool mixing and stratification analysis using GOTHIC lumped parameter modeling methodology

    International Nuclear Information System (INIS)

    Ozdemir, Ozkan Emre; George, Thomas L.

    2015-01-01

    As a part of the GOTHIC (GOTHIC incorporates technology developed for the electric power industry under the sponsorship of EPRI.) Fukushima Technical Evaluation project (EPRI, 2014a, b, 2015), GOTHIC (EPRI, 2014c) has been benchmarked against test data for pool stratification (EPRI, 2014a, b, Ozdemir and George, 2013). These tests confirmed GOTHIC’s ability to simulate pool mixing and stratification under a variety of anticipated suppression pool operating conditions. The multidimensional modeling requires long simulation times for events that may occur over a period of hours or days. For these scenarios a lumped model of the pressure suppression chamber is desirable to maintain reasonable simulation times. However, a lumped model for the pool is not able to predict the effects of pool stratification that can influence the overall containment response. The main objective of this work is on the development of a correlation that can be used to estimate pool mixing and stratification effects in a lumped modeling approach. A simplified lumped GOTHIC model that includes a two zone model for the suppression pool with controlled circulation between the upper and lower zones was constructed. A pump and associated flow connections are included to provide mixing between the upper and lower pool volumes. Using numerically generated data from a multidimensional GOTHIC model for the suppression pool, a correlation was developed for the mixing rate between the upper and lower pool volumes in a two-zone, lumped model. The mixing rate depends on the pool subcooling, the steam injection rate and the injection depth

  18. Simple models for the simulation of submarine melt for a Greenland glacial system model

    Science.gov (United States)

    Beckmann, Johanna; Perrette, Mahé; Ganopolski, Andrey

    2018-01-01

    Two hundred marine-terminating Greenland outlet glaciers deliver more than half of the annually accumulated ice into the ocean and have played an important role in the Greenland ice sheet mass loss observed since the mid-1990s. Submarine melt may play a crucial role in the mass balance and position of the grounding line of these outlet glaciers. As the ocean warms, it is expected that submarine melt will increase, potentially driving outlet glaciers retreat and contributing to sea level rise. Projections of the future contribution of outlet glaciers to sea level rise are hampered by the necessity to use models with extremely high resolution of the order of a few hundred meters. That requirement in not only demanded when modeling outlet glaciers as a stand alone model but also when coupling them with high-resolution 3-D ocean models. In addition, fjord bathymetry data are mostly missing or inaccurate (errors of several hundreds of meters), which questions the benefit of using computationally expensive 3-D models for future predictions. Here we propose an alternative approach built on the use of a computationally efficient simple model of submarine melt based on turbulent plume theory. We show that such a simple model is in reasonable agreement with several available modeling studies. We performed a suite of experiments to analyze sensitivity of these simple models to model parameters and climate characteristics. We found that the computationally cheap plume model demonstrates qualitatively similar behavior as 3-D general circulation models. To match results of the 3-D models in a quantitative manner, a scaling factor of the order of 1 is needed for the plume models. We applied this approach to model submarine melt for six representative Greenland glaciers and found that the application of a line plume can produce submarine melt compatible with observational data. Our results show that the line plume model is more appropriate than the cone plume model for simulating

  19. Models and correlations of the DEBRIS Late-Phase Melt Progression Model

    International Nuclear Information System (INIS)

    Schmidt, R.C.; Gasser, R.D.

    1997-09-01

    The DEBRIS Late Phase Melt Progression Model is an assembly of models, embodied in a computer code, which is designed to treat late-phase melt progression in dry rubble (or debris) regions that can form as a consequence of a severe core uncover accident in a commercial light water nuclear reactor. The approach is fully two-dimensional, and incorporates a porous medium modeling framework together with conservation and constitutive relationships to simulate the time-dependent evolution of such regions as various physical processes act upon the materials. The objective of the code is to accurately model these processes so that the late-phase melt progression that would occur in different hypothetical severe nuclear reactor accidents can be better understood and characterized. In this report the models and correlations incorporated and used within the current version of DEBRIS are described. These include the global conservation equations solved, heat transfer and fission heating models, melting and refreezing models (including material interactions), liquid and solid relocation models, gas flow and pressure field models, and the temperature and compositionally dependent material properties employed. The specific models described here have been used in the experiment design analysis of the Phebus FPT-4 debris-bed fission-product release experiment. An earlier DEBRIS code version was used to analyze the MP-1 and MP-2 late-phase melt progression experiments conducted at Sandia National Laboratories for the US Nuclear Regulatory Commission

  20. Model and simulation for melt flow in micro-injection molding based on the PTT model

    International Nuclear Information System (INIS)

    Cao, Wei; Kong, Lingchao; Li, Qian; Ying, Jin; Shen, Changyu

    2011-01-01

    Unsteady viscoelastic flows were studied using the finite element method in this work. The Phan-Thien–Tanner (PTT) model was used to represent the rheological behavior of viscoelastic fluids. To effectively describe the microscale effects, the slip boundary condition and surface tension were added to the mathematical model for melt flow in micro-injection molding. The new variational equation of pressure, including the viscoelastic parameters and slip boundary condition, was generalized using integration by parts. A computer code based on the finite element method and finite difference method was developed to solve the melt flow problem. Numerical simulation revealed that the melt viscoelasticity plays an important role in the prediction of melt pressure, temperature at the gate and the succeeding melt front advancement in the cavity. Using the viscoelastic model one can also control the rapid increase in simulated pressure, temperature, and reduce the filling difference among different cavities. The short shot experiments of micro-motor shaft showed that the predicted melt front from the viscoelastic model is in fair agreement with the corresponding experimental results

  1. Turbulence model for melt pool natural convection heat transfer

    International Nuclear Information System (INIS)

    Kelkar, K.M.; Patankar, S.V.

    1994-01-01

    Under severe reactor accident scenarios, pools of molten core material may form in the reactor core or in the hemispherically shaped lower plenum of the reactor vessel. Such molten pools are internally heated due to the radioactive decay heat that gives rise to buoyant flows in the molten pool. The flow in such pools is strongly influenced by the turbulent mixing because the expected Rayleigh numbers under accidents scenarios are very high. The variation of the local heat flux over the boundaries of the molten pools are important in determining the subsequent melt progression behavior. This study reports results of an ongoing effort towards providing a well validated mathematical model for the prediction of buoyant flow and heat transfer in internally heated pool under conditions expected in severe accident scenarios

  2. Development of two phase turbulent mixing model for subchannel analysis relevant to BWR

    International Nuclear Information System (INIS)

    Sharma, M.P.; Nayak, A.K.; Kannan, Umasankari

    2014-01-01

    A two phase flow model is presented, which predicts both liquid and gas phase turbulent mixing rate between adjacent subchannels of reactor rod bundles. The model presented here is for slug churn flow regime, which is dominant as compared to the other regimes like bubbly flow and annular flow regimes, since turbulent mixing rate is the highest in slug churn flow regime. In this paper, we have defined new dimensionless parameters i.e. liquid mixing number and gas mixing number for two phase turbulent mixing. The liquid mixing number is a function of mixture Reynolds number whereas the gas phase mixing number is a function of both mixture Reynolds number and volumetric fraction of gas. The effect of pressure, geometrical influence of subchannel is also included in this model. The present model has been tested against low pressure and temperature air-water and high pressure and temperature steam-water experimental data found that it shows good agreement with available experimental data. (author)

  3. Kinetic approach in numerical modeling of melting and crystallization at laser cladding with powder injection

    Energy Technology Data Exchange (ETDEWEB)

    Mirzade, F. Kh., E-mail: fmirzade@rambler.ru [Institute on Laser and Information Technology, Russian Academy of Sciences, 1 Svyatoozerskaya Street, Shatura, Moscow Region 140700 (Russian Federation); Niziev, V.G.; Panchenko, V. Ya.; Khomenko, M.D.; Grishaev, R.V. [Institute on Laser and Information Technology, Russian Academy of Sciences, 1 Svyatoozerskaya Street, Shatura, Moscow Region 140700 (Russian Federation); Pityana, S.; Rooyen, Corney van [CSIR-National Laser Centre, Building 46A, Meiring Nauder Road, Brummeria, Pretoria (South Africa)

    2013-08-15

    The numerical model of laser cladding with coaxial powder injection includes the equations for heat transfer, melting and crystallization kinetics. It has been shown that the main parameters influencing the melt pool dynamics and medium maximum temperature are mass feed rate, laser power and scanning velocity. It has been observed that, due to the phase change occurring with superheating/undercooling, the melt zone has the boundary distinguished from melting isotherm. The calculated melt pool dimensions and dilution are in a good agreement with the experimental results for cladding of 431 martensitic stainless steel onto carbon steel substrate.

  4. Multiphysics modeling of selective laser sintering/melting

    Science.gov (United States)

    Ganeriwala, Rishi Kumar

    A significant percentage of total global employment is due to the manufacturing industry. However, manufacturing also accounts for nearly 20% of total energy usage in the United States according to the EIA. In fact, manufacturing accounted for 90% of industrial energy consumption and 84% of industry carbon dioxide emissions in 2002. Clearly, advances in manufacturing technology and efficiency are necessary to curb emissions and help society as a whole. Additive manufacturing (AM) refers to a relatively recent group of manufacturing technologies whereby one can 3D print parts, which has the potential to significantly reduce waste, reconfigure the supply chain, and generally disrupt the whole manufacturing industry. Selective laser sintering/melting (SLS/SLM) is one type of AM technology with the distinct advantage of being able to 3D print metals and rapidly produce net shape parts with complicated geometries. In SLS/SLM parts are built up layer-by-layer out of powder particles, which are selectively sintered/melted via a laser. However, in order to produce defect-free parts of sufficient strength, the process parameters (laser power, scan speed, layer thickness, powder size, etc.) must be carefully optimized. Obviously, these process parameters will vary depending on material, part geometry, and desired final part characteristics. Running experiments to optimize these parameters is costly, energy intensive, and extremely material specific. Thus a computational model of this process would be highly valuable. In this work a three dimensional, reduced order, coupled discrete element - finite difference model is presented for simulating the deposition and subsequent laser heating of a layer of powder particles sitting on top of a substrate. Validation is provided and parameter studies are conducted showing the ability of this model to help determine appropriate process parameters and an optimal powder size distribution for a given material. Next, thermal stresses upon

  5. The thermo-elastic instability model of melting of alkali halides in the Debye approximation

    Science.gov (United States)

    Owens, Frank J.

    2018-05-01

    The Debye model of lattice vibrations of alkali halides is used to show that there is a temperature below the melting temperature where the vibrational pressure exceeds the electrostatic pressure. The onset temperature of this thermo-elastic instability scales as the melting temperature of NaCl, KCl, and KBr, suggesting its role in the melting of the alkali halides in agreement with a previous more rigorous model.

  6. A model for the latent heat of melting in free standing metal nanoparticles

    International Nuclear Information System (INIS)

    Shin, Jeong-Heon; Deinert, Mark R.

    2014-01-01

    Nanoparticles of many metals are known to exhibit scale dependent latent heats of melting. Analytical models for this phenomenon have so far failed to completely capture the observed phenomena. Here we present a thermodynamic analysis for the melting of metal nanoparticles in terms of their internal energy and a scale dependent surface tension proposed by Tolman. The resulting model predicts the scale dependence of the latent heat of melting and is confirmed using published data for tin and aluminum

  7. Modeling of velocity field for vacuum induction melting process

    Institute of Scientific and Technical Information of China (English)

    CHEN Bo; JIANG Zhi-guo; LIU Kui; LI Yi-yi

    2005-01-01

    The numerical simulation for the recirculating flow of melting of an electromagnetically stirred alloy in a cylindrical induction furnace crucible was presented. Inductive currents and electromagnetic body forces in the alloy under three different solenoid frequencies and three different melting powers were calculated, and then the forces were adopted in the fluid flow equations to simulate the flow of the alloy and the behavior of the free surface. The relationship between the height of the electromagnetic stirring meniscus, melting power, and solenoid frequency was derived based on the law of mass conservation. The results show that the inductive currents and the electromagnetic forces vary with the frequency, melting power, and the physical properties of metal. The velocity and the height of the meniscus increase with the increase of the melting power and the decrease of the solenoid frequency.

  8. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

  9. Experimental Melting Study of Basalt-Peridotite Hybrid Source: Melting model of Hawaiian plume

    Science.gov (United States)

    Takahashi, E.; Gao, S.

    2015-12-01

    Eclogite component entrained in ascending plume is considered to be essentially important in producing flood basalts (e.g., Columbia River basalt, Takahashi et al., 1998 EPSL), alkalic OIBs (e.g., Kogiso et al.,2003), ferro-picrites (Tuff et al.,2005) and Hawaiian shield lavas (e.g., Hauri, 1996; Takahashi & Nakajima, 2002, Sobolev et al.,2005). Size of the entrained eclogite, which controls the reaction rates with ambient peridotite, however, is very difficult to constrain using geophysical observation. Among Hawaiian shield volcanoes, Koolau is the most enriched end-member in eclogite component (Frey et al, 1994). Reconstruction of Koolau volcano based on submarine study on Nuuanu landslide (AGU Monograph vol.128, 2002, Takahashi Garcia Lipman eds.) revealed that silica-rich tholeiite appeared only at the last stage (Makapuu stage) of Koolau volcano. Chemical compositions of lavas as well as isotopes change abruptly and coherently across a horizon (Shinozaki et al. and Tanaka et al. ibid.). Based on these observation, Takahashi & Nakajima (2002 ibid) proposed that the Makapuu stage lava in Koolau volcano was supplied from a single large eclogite block. In order to study melting process in Hawaiian plume, high-pressure melting experiments were carried out under dry and hydrous conditions with layered eclogite/peridotite starting materials. Detail of our experiments will be given by Gao et al (2015 AGU). Combined previous field observation with new set of experiments, we propose that variation in SiO2 among Hawaiian tholeiites represent varying degree of wall-rock interaction between eclogite and ambient peridotite. Makapuu stage lavas in Koolau volcano represents eclogite partial melts formed at ~3 GPa with various amount of xenocrystic olivines derived from Pacific plate. In other words, we propose that "primary magma" in the melting column of Hawaiian plume ranges from basaltic andesite to ferro-picrite depending on the lithology of the source. Solidus of

  10. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    Lopez R, A.

    2004-01-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  11. LOSP-initiated event tree analysis for BWR

    International Nuclear Information System (INIS)

    Watanabe, Norio; Kondo, Masaaki; Uno, Kiyotaka; Chigusa, Takeshi; Harami, Taikan

    1989-03-01

    As a preliminary study of 'Japanese Model Plant PSA', a LOSP (loss of off-site power)-initiated Event Tree Analysis for a Japanese typical BWR was carried out solely based on the open documents such as 'Safety Analysis Report'. The objectives of this analysis are as follows; - to delineate core-melt accident sequences initiated by LOSP, - to evaluate the importance of core-melt accident sequences in terms of occurrence frequency, and - to develop a foundation of plant information and analytical procedures for efficiently performing further 'Japanese Model Plant PSA'. This report describes the procedure and results of the LOSP-initiated Event Tree Analysis. In this analysis, two types of event trees, Functional Event Tree and Systemic Event Tree, were developed to delineate core-melt accident sequences and to quantify their frequencies. Front-line System Event Tree was prepared as well to provide core-melt sequence delineation for accident progression analysis of Level 2 PSA which will be followed in a future. Applying U.S. operational experience data such as component failure rates and a LOSP frequency, we obtained the following results; - The total frequency of core-melt accident sequences initiated by LOSP is estimated at 5 x 10 -4 per reactor-year. - The dominant sequences are 'Loss of Decay Heat Removal' and 'Loss of Emergency Electric Power Supply', which account for more than 90% of the total core-melt frequency. In this analysis, a higher value of 0.13/R·Y was used for the LOSP frequency than experiences in Japan and any recovery action was not considered. In fact, however, there has been no experience of LOSP event in Japanese nuclear power plants so far and it is also expected that offsite power and/or PCS would be recovered before core melt. Considering Japanese operating experience and recovery factors will reduce the total core-melt frequency to less than 10 -6 per reactor-year. (J.P.N.)

  12. Computationally efficient thermal-mechanical modelling of selective laser melting

    Science.gov (United States)

    Yang, Yabin; Ayas, Can

    2017-10-01

    The Selective laser melting (SLM) is a powder based additive manufacturing (AM) method to produce high density metal parts with complex topology. However, part distortions and accompanying residual stresses deteriorates the mechanical reliability of SLM products. Modelling of the SLM process is anticipated to be instrumental for understanding and predicting the development of residual stress field during the build process. However, SLM process modelling requires determination of the heat transients within the part being built which is coupled to a mechanical boundary value problem to calculate displacement and residual stress fields. Thermal models associated with SLM are typically complex and computationally demanding. In this paper, we present a simple semi-analytical thermal-mechanical model, developed for SLM that represents the effect of laser scanning vectors with line heat sources. The temperature field within the part being build is attained by superposition of temperature field associated with line heat sources in a semi-infinite medium and a complimentary temperature field which accounts for the actual boundary conditions. An analytical solution of a line heat source in a semi-infinite medium is first described followed by the numerical procedure used for finding the complimentary temperature field. This analytical description of the line heat sources is able to capture the steep temperature gradients in the vicinity of the laser spot which is typically tens of micrometers. In turn, semi-analytical thermal model allows for having a relatively coarse discretisation of the complimentary temperature field. The temperature history determined is used to calculate the thermal strain induced on the SLM part. Finally, a mechanical model governed by elastic-plastic constitutive rule having isotropic hardening is used to predict the residual stresses.

  13. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

  14. Numerical modelling of an industrial glass-melting furnace

    Energy Technology Data Exchange (ETDEWEB)

    Hill, S C [Brigham Young Univ., Advanced Combustion Engineering Research Center, Provo, UT (United States); Webb, B W; McQuay, M Q [Brigham Young Univ., Mechanical Engineering Dept., Provo, UT (United States); Newbold, J [Lockheed Aerospace, Denver, CO (United States)

    2000-03-01

    The predictive capability of two comprehensive combustion codes, PCGC-3 and FLUENT, to simulate local flame structure and combustion characteristics in a industrial gas-fired, flat-glass furnace is investigated. Model predictions are compared with experimental data from the furnace for profiles of velocity, species concentrations, temperatures, and wall-incident radiative heat flux. Predictions from both codes show agreement with the measured mean velocity profiles and incident radiant flux on the crown. However, significant differences between the code predictions and measurements are observed for the flame-ozone temperatures and species concentrations. The observed discrepancies may be explained by (i) uncertainties in the distributions of mean velocity and turbulence in the portneck, (ii) uncertainties in the port-by-port stoichiometry, (iii) different grid-based approximations to the furnace geometry made in the two codes, (iv) the assumption of infinitely fast chemistry made in the chemical reaction model of both codes, and (v) simplifying assumptions made in the simulations regarding the complex coupling between the combustion space, batch blanket, and melt tank. The study illustrates the critical need for accurate boundary conditions (inlet air and fuel flow distributions, boundary surface temperatures, etc.) and the importance of representative furnace geometry in simulating these complex industrial combustion systems. (Author)

  15. Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

  16. Modeling the impact of melt on seismic properties during mountain building

    Science.gov (United States)

    Lee, Amicia L.; Walker, Andrew M.; Lloyd, Geoffrey E.; Torvela, Taija

    2017-03-01

    Initiation of partial melting in the mid/lower crust causes a decrease in P wave and S wave velocities; recent studies imply that the relationship between these velocities and melt is not simple. We have developed a modeling approach to assess the combined impact of various melt and solid phase properties on seismic velocities and anisotropy. The modeling is based on crystallographic preferred orientation (CPO) data measured from migmatite samples, allowing quantification of the variation of seismic velocities with varying melt volumes, shapes, orientations, and matrix anisotropy. The results show nonlinear behavior of seismic properties as a result of the interaction of all of these physical properties, which in turn depend on lithology, stress regime, strain rate, preexisting rock fabrics, and pressure-temperature conditions. This nonlinear behavior is evident when applied to a suite of samples from a traverse across a migmatitic shear zone in the Seiland Igneous Province, Northern Norway. Critically, changes in solid phase composition and CPO, and melt shape and orientation with respect to the wave propagation direction can result in huge variations in the same seismic property even if the melt fraction remains the same. A comparison with surface wave interpretations from tectonically active regions highlights the issues in current models used to predict melt percentages or partially molten regions. Interpretation of seismic data to infer melt percentages or extent of melting should, therefore, always be underpinned by robust modeling of the underlying geological parameters combined with examination of multiple seismic properties in order to reduce uncertainty of the interpretation.

  17. A computational model for viscous fluid flow, heat transfer, and melting in in situ vitrification melt pools

    International Nuclear Information System (INIS)

    McHugh, P.R.; Ramshaw, J.D.

    1991-11-01

    MAGMA is a FORTRAN computer code designed to viscous flow in in situ vitrification melt pools. It models three-dimensional, incompressible, viscous flow and heat transfer. The momentum equation is coupled to the temperature field through the buoyancy force terms arising from the Boussinesq approximation. All fluid properties, except density, are assumed variable. Density is assumed constant except in the buoyancy force terms in the momentum equation. A simple melting model based on the enthalpy method allows the study of the melt front progression and latent heat effects. An indirect addressing scheme used in the numerical solution of the momentum equation voids unnecessary calculations in cells devoid of liquid. Two-dimensional calculations can be performed using either rectangular or cylindrical coordinates, while three-dimensional calculations use rectangular coordinates. All derivatives are approximated by finite differences. The incompressible Navier-Stokes equations are solved using a new fully implicit iterative technique, while the energy equation is differenced explicitly in time. Spatial derivatives are written in conservative form using a uniform, rectangular, staggered mesh based on the marker and cell placement of variables. Convective terms are differenced using a weighted average of centered and donor cell differencing to ensure numerical stability. Complete descriptions of MAGMA governing equations, numerics, code structure, and code verification are provided. 14 refs

  18. A computational model for viscous fluid flow, heat transfer, and melting in in situ vitrification melt pools

    Energy Technology Data Exchange (ETDEWEB)

    McHugh, P.R.; Ramshaw, J.D.

    1991-11-01

    MAGMA is a FORTRAN computer code designed to viscous flow in in situ vitrification melt pools. It models three-dimensional, incompressible, viscous flow and heat transfer. The momentum equation is coupled to the temperature field through the buoyancy force terms arising from the Boussinesq approximation. All fluid properties, except density, are assumed variable. Density is assumed constant except in the buoyancy force terms in the momentum equation. A simple melting model based on the enthalpy method allows the study of the melt front progression and latent heat effects. An indirect addressing scheme used in the numerical solution of the momentum equation voids unnecessary calculations in cells devoid of liquid. Two-dimensional calculations can be performed using either rectangular or cylindrical coordinates, while three-dimensional calculations use rectangular coordinates. All derivatives are approximated by finite differences. The incompressible Navier-Stokes equations are solved using a new fully implicit iterative technique, while the energy equation is differenced explicitly in time. Spatial derivatives are written in conservative form using a uniform, rectangular, staggered mesh based on the marker and cell placement of variables. Convective terms are differenced using a weighted average of centered and donor cell differencing to ensure numerical stability. Complete descriptions of MAGMA governing equations, numerics, code structure, and code verification are provided. 14 refs.

  19. Kinematics of two-phase mixture level motion in BWR pressure vessels

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

    1985-01-01

    A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

  20. Mathematical modeling of quartz particle melting process in plasma-chemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Volokitin, Oleg, E-mail: volokitin-oleg@mail.ru; Volokitin, Gennady, E-mail: vgg-tomsk@mail.ru; Skripnikova, Nelli, E-mail: nks2003@mai.ru; Shekhovtsov, Valentin, E-mail: shehovcov2010@yandex.ru [Tomsk State University of Architecture and Building, 2, Solyanaya Sq., 634003, Tomsk (Russian Federation); Vlasov, Viktor, E-mail: rector@tsuab.ru [Tomsk State University of Architecture and Building, 2, Solyanaya Sq., 634003, Tomsk (Russian Federation); National Research Tomsk Polytechnic University, 30, Lenin Ave., 634050, Tomsk (Russian Federation)

    2016-01-15

    Among silica-based materials vitreous silica has a special place. The paper presents the melting process of a quartz particle under conditions of low-temperature plasma. A mathematical model is designed for stages of melting in the experimental plasma-chemical reactor. As calculation data show, quartz particles having the radius of 0.21≤ r{sub p} ≤0.64 mm completely melt at W = 0.65 l/s particle feed rate depending on the Nusselt number, while 0.14≤ r{sub p} ≤0.44 mm particles melt at W = 1.4 l/s. Calculation data showed that 2 mm and 0.4 mm quartz particles completely melted during and 0.1 s respectively. Thus, phase transformations occurred in silicon dioxide play the important part in its heating up to the melting temperature.

  1. Benchmarking of AREVA BWR FDIC-PEZOG model against first BFE3 cycle 15 application of On-Line NobleChem results

    International Nuclear Information System (INIS)

    Pop, M.G.; Lamanna, L.S.; Hoornik, A.; Storey, G.C.; Lemons, J.F.

    2015-01-01

    The combination of AREVA's BWR FDIC-PEZOG tools allows the calculation of the total liftoff as a measure of fuel performance and a risk indicator for fuel reliability. The AREVA BWR FDIC tool is a crud modeling tool. The PEZOG tool models the platinum-enhanced zirconium oxide growth of fuel cladding when exposed to platinum during operation. Continuous effort to improve these tools used for the total liftoff calculations is illustrated by the benchmarking of the tools after the application of On-Line NobleChem TM at TVA Browns Ferry Unit 3 during Cycle 15. A set of runs using the modified FDIC-PEZOG model and actual plant water chemistry for Cycle 15 and partial data for Cycle 16 were performed. The updated results' deposit thickness and deposit composition predictions for EOC15 were compared to the measured data from EOC15 and are presented in this paper. The updated predicted deposit thickness matched the actual, measured value exactly. Predicted deposit composition near the fuel rod boundary, nearer to the bulk reactor water, and as an averaged deposit, as presented in the paper, compared extremely well with the measured data at EOC15. The updated AREVA methodology resulted in lower fuel oxide thickness predictions over the life of the fuel as compared to the initial evaluations for BFE3 by incorporating more recent experimental data on the thermal conductivity of zirconia; unnecessary conservatism in the prediction of the fuel oxide thickness over the life of the fuel was removed in the improved model. (authors)

  2. Two phase modeling of nanofluid flow in existence of melting heat transfer by means of HAM

    Science.gov (United States)

    Sheikholeslami, M.; Jafaryar, M.; Bateni, K.; Ganji, D. D.

    2018-02-01

    In this article, Buongiorno Model is applied for investigation of nanofluid flow over a stretching plate in existence of magnetic field. Radiation and Melting heat transfer are taken into account. Homotopy analysis method (HAM) is selected to solve ODEs which are obtained from similarity transformation. Roles of Brownian motion, thermophoretic parameter, Hartmann number, porosity parameter, Melting parameter and Eckert number are presented graphically. Results indicate that nanofluid velocity and concentration enhance with rise of melting parameter. Nusselt number reduces with increase of porosity and melting parameters.

  3. BWR stability analysis

    International Nuclear Information System (INIS)

    Valtonen, K.

    1990-01-01

    The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

  4. Residual stress analysis in BWR pressure vessel attachments

    International Nuclear Information System (INIS)

    Dexter, R.J.; Leung, C.P.; Pont, D.

    1992-06-01

    Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

  5. Development of a BWR core burn-up calculation code COREBN-BWR

    International Nuclear Information System (INIS)

    Morimoto, Yuichi; Okumura, Keisuke

    1992-05-01

    In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

  6. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

  7. Pattern recognition model to estimate intergranular stress corrosion cracking (IGSCC) at crevices and pit sites of 304 SS in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Urquidi-Macdonald, Mirna [Penn State University, 212 Earth-Engineering Science Building, University Park, PA 16801 (United States)

    2004-07-01

    Many publications have shown that crack growth rates (CGR) due to intergranular stress corrosion cracking (IGSCC) of metals is dependent on many parameters related to the manufacturing process of the steel and the environment to which the steel is exposed. Those parameters include, but are not restricted to, the concentration of chloride, fluoride, nitrates, and sulfates, pH, fluid velocity, electrochemical potential (ECP), electrolyte conductivity, stress and sensitization applied to the steel during its production and use. It is not well established how combinations of each of these parameters impact the CGR. Many different models and beliefs have been published, resulting in predictions that sometimes disagree with experimental observations. To some extent, the models are the closest to the nature of IGSCC, however, there is not a model that fully describes the entire range of observations, due to the difficulty of the problem. Among the models, the Fracture Environment Model, developed by Macdonald et al., is the most physico-chemical model, accounting for experimental observations in a wide range of environments or ECPs. In this work, we collected experimental data on BWR environments and designed a data mining pattern recognition model to learn from that data. The model was used to generate CGR estimations as a function of ECP on a BWR environment. The results of the predictive model were compared to the Fracture Environment Model predictions. The results from those two models are very close to the experimental observations of the area corresponding to creep and IGSCC controlled by diffusion. At more negative ECPs than the potential corresponding to creep, the pattern recognition predicts an increase of CGR with decreasing ECP, while the Fracture Environment Model predicts the opposite. The results of this comparison confirm that the pattern recognition model covers 3 phenomena: hydrogen embrittlement at very negative ECP, creep at intermediate ECP, and IGSCC

  8. Pattern recognition model to estimate intergranular stress corrosion cracking (IGSCC) at crevices and pit sites of 304 SS in BWR environments

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, Mirna

    2004-01-01

    Many publications have shown that crack growth rates (CGR) due to intergranular stress corrosion cracking (IGSCC) of metals is dependent on many parameters related to the manufacturing process of the steel and the environment to which the steel is exposed. Those parameters include, but are not restricted to, the concentration of chloride, fluoride, nitrates, and sulfates, pH, fluid velocity, electrochemical potential (ECP), electrolyte conductivity, stress and sensitization applied to the steel during its production and use. It is not well established how combinations of each of these parameters impact the CGR. Many different models and beliefs have been published, resulting in predictions that sometimes disagree with experimental observations. To some extent, the models are the closest to the nature of IGSCC, however, there is not a model that fully describes the entire range of observations, due to the difficulty of the problem. Among the models, the Fracture Environment Model, developed by Macdonald et al., is the most physico-chemical model, accounting for experimental observations in a wide range of environments or ECPs. In this work, we collected experimental data on BWR environments and designed a data mining pattern recognition model to learn from that data. The model was used to generate CGR estimations as a function of ECP on a BWR environment. The results of the predictive model were compared to the Fracture Environment Model predictions. The results from those two models are very close to the experimental observations of the area corresponding to creep and IGSCC controlled by diffusion. At more negative ECPs than the potential corresponding to creep, the pattern recognition predicts an increase of CGR with decreasing ECP, while the Fracture Environment Model predicts the opposite. The results of this comparison confirm that the pattern recognition model covers 3 phenomena: hydrogen embrittlement at very negative ECP, creep at intermediate ECP, and IGSCC

  9. Structural test and analysis of a model of a BWR suppression chamber support in the plastic regime

    International Nuclear Information System (INIS)

    Blumer, U.R.; Klaeui, E.; Bosshard, E.P.

    1991-01-01

    A BWR Mark I suppression pool support has been analysed and tested in the laboratory. The aim was the demonstration of the acceptability of hypothetical dynamic loadings resulting from simultaneous steam blowdown through all safety relief valves. The analysis has shown that plastic deformation will locally occur, which is difficult to assess purely theoretical. Therefore tests in reduced scale were performed that show the amount and distribution of plastic flow in the supports. The paper describes the elastic analysis, the theory of the scaling laws for the reduced scale test, the test and its results. It also shows the thermographical method that has been used to determine the plastic material flow in the support structure. (author)

  10. Heat transfer model and finite element formulation for simulation of selective laser melting

    Science.gov (United States)

    Roy, Souvik; Juha, Mario; Shephard, Mark S.; Maniatty, Antoinette M.

    2017-10-01

    A novel approach and finite element formulation for modeling the melting, consolidation, and re-solidification process that occurs in selective laser melting additive manufacturing is presented. Two state variables are introduced to track the phase (melt/solid) and the degree of consolidation (powder/fully dense). The effect of the consolidation on the absorption of the laser energy into the material as it transforms from a porous powder to a dense melt is considered. A Lagrangian finite element formulation, which solves the governing equations on the unconsolidated reference configuration is derived, which naturally considers the effect of the changing geometry as the powder melts without needing to update the simulation domain. The finite element model is implemented into a general-purpose parallel finite element solver. Results are presented comparing to experimental results in the literature for a single laser track with good agreement. Predictions for a spiral laser pattern are also shown.

  11. Melt pool modelling, simulation and experimental validation for SLM

    NARCIS (Netherlands)

    Wits, Wessel

    2017-01-01

    SLM parts are built by successively melting layers of powder in a powder bed. Process parameters are often optimized experimentally by laser scanning a number of single tracks and subsequently determining which settings lead to a good compromise between quality and build speed. However,

  12. In-BWR and out-of-pile nodular corrosion behavior of Zry-2/4 type melts with varying Fe, Cr, and Ni content and varying process history

    International Nuclear Information System (INIS)

    Ruhmann, H.; Manzel, R.; Sell, H.J.

    1996-01-01

    Zircaloy-based materials with constant tin content, a constant sum of Fe and Cr content, and different Fe/Cr ratios were manufactured from small ingots (6 kg) by forging, β-quenching, hot rolling, and cold rolling with two different annealing sequences resulting in two accumulated annealing parameters. To study the effect of Ni, chromium was substituted by nickel in one alloy. As a reference material standard, ASTM Zry-2 was manufactured in a similar way. These materials were examined for precipitate size by TEM and for their out-of-pile corrosion behavior in 400 C steam up to 319 days and in 500 C steam up to 24 h in a static and in a refreshed autoclave. Furthermore, samples made from these alloys were irradiated in a commercial BWR for two years. Results of the test are given

  13. Modeling of melt retention in EU-APR1400 ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V. S.; Sulatsky, A. A.; Khabensky, V. B.; Sulatskaya, M. B. [Alexandrov Research Inst. of Technology NITI, Sosnovy Bor (Russian Federation); Gusarov, V. V.; Almyashev, V. I.; Komlev, A. A. [Saint Petersburg State Technological Univ. SPbSTU, St.Petersburg (Russian Federation); Bechta, S. [KTH, Stockholm (Sweden); Kim, Y. S. [KHNP, 1312 Gil 70, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Park, R. J.; Kim, H. Y.; Song, J. H. [KAERI, 989 Gil 111, Daedeokdaero, Yuseong-gu, Daejeon (Korea, Republic of)

    2012-07-01

    A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)

  14. Modeling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Bui, V.A.; Nourgaliev, R.R. [Royal Institute of Technology, Stockholm (Sweden)] [and others

    1995-09-01

    The objective of the paper is to study heat and mass transfer processes related to core melt discharge from a reactor vessel is a severe light water reactor accident. The phenomenology of the issue includes (1) melt convection in and heat transfer from the melt pool in contact with the vessel lower head wall; (2) fluid dynamics and heat transfer of the melt flow in the growing discharge hole; and (3) multi-dimensional heat conduction in the ablating lower head wall. A program of model development, validation and application is underway (i) to analyse the dominant physical mechanisms determining characteristics of the lower head ablation process; (ii) to develop and validate efficient analytic/computational methods for estimating heat and mass transfer under phase-change conditions in irregular moving-boundary domains; and (iii) to investigate numerically the melt discharge phenomena in a reactor-scale situation, and, in particular, the sensitivity of the melt discharge transient to structural differences and various in-vessel melt progression scenarios. The paper presents recent results of the analysis and model development work supporting the simulant melt-structure interaction experiments.

  15. Two-dimensional model of laser alloying of binary alloy powder with interval of melting temperature

    Science.gov (United States)

    Knyzeva, A. G.; Sharkeev, Yu. P.

    2017-10-01

    The paper contains two-dimensional model of laser beam melting of powders from binary alloy. The model takes into consideration the melting of alloy in some temperature interval between solidus and liquidus temperatures. The external source corresponds to laser beam with energy density distributed by Gauss law. The source moves along the treated surface according to given trajectory. The model allows investigating the temperature distribution and thickness of powder layer depending on technological parameters.

  16. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

    Garner, G.; Divakaruni, S.M.; Meyer, J.E.

    1986-01-01

    Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

  17. Three-dimensional model of heat transport during In Situ Vitrification with melting and cool down

    International Nuclear Information System (INIS)

    Hawkes, G.L.

    1993-01-01

    A potential technology for permanent remediation of buried wastes is the In Situ Vitrification (ISV) process. This process uses electrical resistance heating to melt waste and contaminated soil in place to produce a durable, glasslike material that encapsulates and immobilizes buried wastes. The magnitude of the resulting electrical resistance heating is sufficient to cause soil melting. As the molten region grows, surface heat losses cause the soil near the surface to re solidify. This paper presents numerical results obtained by considering heat transport and melting when solving the conservation of mass and energy equations using finite element methods. A local heat source is calculated by solving the electric field equation and calculating a Joule Heat source term. The model considered is a three-dimensional model of the electrodes and surrounding soil. Also included in the model is subsidence; where the surface of the melted soil subsides due to the change in density when the soil melts. A power vs. time profile is implemented for typical ISV experiments. The model agrees well with experimental data for melt volume and melt shape

  18. Best-estimate analysis development for BWR systems

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

    1986-01-01

    The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

  19. Melt/concrete interactions: the Sandia experimental program, model development, and code comparison test

    International Nuclear Information System (INIS)

    Powers, D.A.; Muir, J.F.

    1979-01-01

    High temperature melt/concrete interactions have been studied both experimentally and analytically at Sandia under sponsorship of Reactor Safety Research of the US Nuclear Regulatory Commission. The purpose of these studies has been to develop an understanding of these interactions suitable for risk assessment. Results of the experimental program are summarized and a computer model of melt/concrete interactions is described. A melt/concrete interaction test that will allow this and other models of the interaction to be compared is also described

  20. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  1. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  2. Development of boiling transition analysis code TCAPE-INS/B based on mechanistic methods for BWR fuel bundles. Models and validations with boiling transition experimental data

    International Nuclear Information System (INIS)

    Ishida, Naoyuki; Utsuno, Hideaki; Kasahara, Fumio

    2003-01-01

    The Boiling Transition (BT) analysis code TCAPE-INS/B based on the mechanistic methods coupled with subchannel analysis has been developed for the evaluation of the integrity of Boiling Water Reactor (BWR) fuel rod bundles under abnormal operations. Objective of the development is the evaluation of the BT without using empirical BT and rewetting correlations needed for different bundle designs in the current analysis methods. TCAPE-INS/B consisted mainly of the drift-flux model, the film flow model, the cross-flow model, the thermal conductivity model and the heat transfer correlations. These models were validated systematically with the experimental data. The accuracy of the prediction for the steady-state Critical Heat Flux (CHF) and the transient temperature of the fuel rod surface after the occurrence of BT were evaluated on the validations. The calculations for the experiments with the single tube and bundles were carried out for the validations of the models incorporated in the code. The results showed that the steady-state CHF was predicted within about 6% average error. In the transient calculations, BT timing and temperature of the fuel rod surface gradient agreed well with experimental results, but rewetting was predicted lately. So, modeling of heat transfer phenomena during post-BT is under modification. (author)

  3. Development of drift-flux model based on 8 x 8 BWR rod bundle geometry experiments under prototypic temperature and pressure conditions

    International Nuclear Information System (INIS)

    Ozaki, Tetsuhiro; Suzuki, Riichiro; Mashiko, Hiroyuki; Hibiki, Takashi

    2013-01-01

    The drift-flux model is one of the imperative concepts used to consider the effects of phase coupling on two-phase flow dynamics. Several drift-flux models are available that apply to rod bundle geometries and some of these are implemented in several nuclear safety analysis codes. However, these models are not validated by well-designed prototypic full bundle test data, and therefore, the scalability of these models has not necessarily been verified. The Nuclear Power Engineering Corporation (NUPEC) conducted void fraction measurement tests in Japan with prototypic 8 x 8 BWR (boiling water reactor) rod bundles under prototypic temperature and pressure conditions. Based on these NUPEC data, a new drift-flux model applicable to predicting the void fraction in a rod bundle geometry has been developed. The newly developed drift-flux model is compared with the other existing data such as the two-phase flow test facility (TPTF) data taken at the Japan Atomic Energy Research Institute (JAERI) [currently, Japan Atomic Energy Agency (JAEA)] and low pressure adiabatic 8 x 8 bundle test data taken at Purdue University in the United States. The results of these comparisons show good agreement between the test data and the predictions. The effects of power distribution, spacer grids, and the bundle geometry on the newly developed drift-flux model have been discussed using the NUPEC data. (author)

  4. Model of coordination melting of crystals and anisotropy of physical and chemical properties of the surface

    Science.gov (United States)

    Bokarev, Valery P.; Krasnikov, Gennady Ya

    2018-02-01

    Based on the evaluation of the properties of crystals, such as surface energy and its anisotropy, the surface melting temperature, the anisotropy of the work function of the electron, and the anisotropy of adsorption, were shown the advantages of the model of coordination melting (MCM) in calculating the surface properties of crystals. The model of coordination melting makes it possible to calculate with an acceptable accuracy the specific surface energy of the crystals, the anisotropy of the surface energy, the habit of the natural crystals, the temperature of surface melting of the crystal, the anisotropy of the electron work function and the anisotropy of the adhesive properties of single-crystal surfaces. The advantage of our model is the simplicity of evaluating the surface properties of the crystal based on the data given in the reference literature. In this case, there is no need for a complex mathematical tool, which is used in calculations using quantum chemistry or modeling by molecular dynamics.

  5. Phase behavior and reactive transport of partial melt in heterogeneous mantle model

    Science.gov (United States)

    Jordan, J.; Hesse, M. A.

    2013-12-01

    The reactive transport of partial melt is the key process that leads to the chemical and physical differentiation of terrestrial planets and smaller celestial bodies. The essential role of the lithological heterogeneities during partial melting of the mantle is increasingly recognized. How far can enriched melts propagate while interacting with the ambient mantle? Can the melt flow emanating from a fertile heterogeneity be localized through a reactive infiltration feedback in a model without exogenous factors or contrived initial conditions? A full understanding of the role of heterogeneities requires reactive melt transport models that account for the phase behavior of major elements. Previous work on reactive transport in the mantle focuses on trace element partitioning; we present the first nonlinear chromatographic analysis of reactive melt transport in systems with binary solid solution. Our analysis shows that reactive melt transport in systems with binary solid solution leads to the formation of two separate reaction fronts: a slow melting/freezing front along which enthalpy change is dominant and a fast dissolution/precipitation front along which compositional changes are dominated by an ion-exchange process over enthalpy change. An intermediate state forms between these two fronts with a bulk-rock composition and enthalpy that are not necessarily bounded by the bulk-rock composition and enthalpy of either the enriched heterogeneity or the depleted ambient mantle. The formation of this intermediate state makes it difficult to anticipate the porosity changes and hence the stability of reaction fronts. Therefore, we develop a graphical representation for the solution that allows identification of the intermediate state by inspection, for all possible bulk-rock compositions and enthalpies of the heterogeneity and the ambient mantle. We apply the analysis to the partial melting of an enriched heterogeneity. This leads to the formation of moving precipitation

  6. Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1992-03-01

    The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

  7. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  8. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  9. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

    2014-01-01

    Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

  10. Melting spectral functions of the scalar and vector mesons in a holographic QCD model

    International Nuclear Information System (INIS)

    Fujita, Mitsutoshi; Kikuchi, Toru; Fukushima, Kenji; Misumi, Tatsuhiro; Murata, Masaki

    2010-01-01

    We investigate the finite-temperature spectral functions of heavy quarkonia by using the soft-wall anti-de Sitter/QCD model. We discuss the scalar, the pseudoscalar, the vector, and the axial-vector mesons and compare their qualitative features of the melting temperature and growing width. We find that the axial-vector meson melts earlier than the vector meson, while there appears only a slight difference between the scalar and pseudoscalar mesons, which also melt earlier than the vector meson.

  11. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  12. Infrared laser-induced chaos and conformational disorder in a model polymer crystal: Melting vs ablation

    International Nuclear Information System (INIS)

    Sumpter, B.G.; Noid, D.W.; Voth, G.A.; Wunderlich, B.

    1990-01-01

    Molecular dynamics-based computer simulations are presented for the interaction of one and two infrared (IR) laser beams with a model polymer surface. When a single laser beam system is studied over a wide range of intensities, only melting of the polymer, or melting followed by bond dissociation, is observed for up to 100 picoseconds. In contrast, the two-laser simulation results exhibit a marked difference in the energy absorption behavior of the irradiated polymer which, in turn, results in multiple bond dissociations. The results for the one- and two-laser cases studied can be divided into four different classes of physical behavior: (a) the polymer remains in the solid state; (b) the polymer crystal melts; (c) the polymer ablates, but with significant melting (charring); or (d) the polymer ablates with minimal melting. Damage to the model polymer crystal from absorption of energy from either one or two lasers occurs through a mechanism that involves the competition between the absorption of energy and internal energy redistribution. The rate of energy loss from the absorption site(s) relative to the rate of absorption of energy from the radiation field determines rather the polymer melts or ablates (low absorption rates lead to melting or no change and high rates lead to ablation). A sufficiently large rate of energy absorption is only obtainable through the use of two lasers. Two lasers also significantly decrease the total laser intensity required to cause polymer crystal melting. The differences between the one- and two-laser cases are studied by adapting novel signal/subspace techniques to analyze the dynamical changes in the mode spectrum of the polymer as it melts

  13. A Dynamic Mesh-Based Approach to Model Melting and Shape of an ESR Electrode

    Science.gov (United States)

    Karimi-Sibaki, E.; Kharicha, A.; Bohacek, J.; Wu, M.; Ludwig, A.

    2015-10-01

    This paper presents a numerical method to investigate the shape of tip and melt rate of an electrode during electroslag remelting process. The interactions between flow, temperature, and electromagnetic fields are taken into account. A dynamic mesh-based approach is employed to model the dynamic formation of the shape of electrode tip. The effect of slag properties such as thermal and electrical conductivities on the melt rate and electrode immersion depth is discussed. The thermal conductivity of slag has a dominant influence on the heat transfer in the system, hence on melt rate of electrode. The melt rate decreases with increasing thermal conductivity of slag. The electrical conductivity of slag governs the electric current path that in turn influences flow and temperature fields. The melting of electrode is a quite unstable process due to the complex interaction between the melt rate, immersion depth, and shape of electrode tip. Therefore, a numerical adaptation of electrode position in the slag has been implemented in order to achieve steady state melting. In fact, the melt rate, immersion depth, and shape of electrode tip are interdependent parameters of process. The generated power in the system is found to be dependent on both immersion depth and shape of electrode tip. In other words, the same amount of power was generated for the systems where the shapes of tip and immersion depth were different. Furthermore, it was observed that the shape of electrode tip is very similar for the systems running with the same ratio of power generation to melt rate. Comparison between simulations and experimental results was made to verify the numerical model.

  14. BWR mechanics and materials technology update

    International Nuclear Information System (INIS)

    Kiss, E.

    1983-01-01

    This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration. (orig.)

  15. Trace Elements in Basalts From the Siqueiros Fracture Zone: Implications for Melt Migration Models

    Science.gov (United States)

    Pickle, R. C.; Forsyth, D. W.; Saal, A. E.; Nagle, A. N.; Perfit, M. R.

    2008-12-01

    Incompatible trace element (ITE) ratios in MORB from a variety of locations may provide insights into the melt migration process by constraining aggregated melt compositions predicted by mantle melting and flow models. By using actual plate geometries to create a 3-D thermodynamic mantle model, melt volumes and compositions at all depths and locations may be calculated and binned into cubes using the pHMELTS algorithm [Asimow et al., 2004]. These melts can be traced from each cube to the surface assuming several migration models, including a simplified pressure gradient model and one in which melt is guided upwards by a low permeability compacted layer. The ITE ratios of all melts arriving at the surface are summed, averaged, and compared to those of the actual sample compositions from the various MOR locales. The Siqueiros fracture zone at 8° 20' N on the East Pacific Rise (EPR) comprises 4 intra-transform spreading centers (ITSCs) across 140 km of offset between two longer spreading ridges, and is an excellent study region for several reasons. First, an abundance of MORB data is readily available, and the samples retrieved from ITSCs are unlikely to be aggregated in a long-lived magma chamber or affected by along-axis transport, so they represent melts extracted locally from the mantle. Additionally, samples at Siqueiros span a compositional range from depleted to normal MORB within the fracture zone yet have similar isotopic compositions to samples collected from the 9-10° EPR. This minimizes the effect of assuming a uniform source composition in our melting model despite a heterogeneous mantle, allowing us to consistently compare the actual lava composition with that predicted by our model. Finally, it has been demonstrated with preliminary migration models that incipient melts generated directly below an ITSC may not necessarily erupt at that ITSC but migrate laterally towards a nearby ridge due to enhanced pressure gradients. The close proximity of the

  16. Using mathematical modeling to control topographical properties of poly (ε-caprolactone) melt electrospun scaffolds

    International Nuclear Information System (INIS)

    Ko, J; Bhullar, S K; Mohtaram, N K; Willerth, S M; Jun, M B G

    2014-01-01

    Melt electrospinning creates fibrous scaffolds using direct deposition. The main challenge of melt electrospinning is controlling the topography of the scaffolds for tissue engineering applications. Mathematical modeling enables a better understanding of the parameters that determine the topography of scaffolds. The objective of this study is to build two types of mathematical models. First, we modeled the melt electrospinning process by incorporating parameters such as nozzle size, counter electrode distance and applied voltage that influence fiber diameter and scaffold porosity. Our second model describes the accumulation of the extruded microfibers on flat and round surfaces using data from the microfiber modeling. These models were validated through the use of experimentally obtained data. Scanning electron microscopy (SEM) was used to image the scaffolds and the fiber diameters were measured using Quartz-PCI Image Management Systems® in SEM to measure scaffold porosity. (paper)

  17. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    Di Auria, F.; Pellicoro, V.

    1995-01-01

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  18. The HAMBO BWR simulator of HAMMLAB

    International Nuclear Information System (INIS)

    Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

    2001-02-01

    Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

  19. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  20. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  1. A simple model for the evolution of melt pond coverage on permeable Arctic sea ice

    Science.gov (United States)

    Popović, Predrag; Abbot, Dorian

    2017-05-01

    As the melt season progresses, sea ice in the Arctic often becomes permeable enough to allow for nearly complete drainage of meltwater that has collected on the ice surface. Melt ponds that remain after drainage are hydraulically connected to the ocean and correspond to regions of sea ice whose surface is below sea level. We present a simple model for the evolution of melt pond coverage on such permeable sea ice floes in which we allow for spatially varying ice melt rates and assume the whole floe is in hydrostatic balance. The model is represented by two simple ordinary differential equations, where the rate of change of pond coverage depends on the pond coverage. All the physical parameters of the system are summarized by four strengths that control the relative importance of the terms in the equations. The model both fits observations and allows us to understand the behavior of melt ponds in a way that is often not possible with more complex models. Examples of insights we can gain from the model are that (1) the pond growth rate is more sensitive to changes in bare sea ice albedo than changes in pond albedo, (2) ponds grow slower on smoother ice, and (3) ponds respond strongest to freeboard sinking on first-year ice and sidewall melting on multiyear ice. We also show that under a global warming scenario, pond coverage would increase, decreasing the overall ice albedo and leading to ice thinning that is likely comparable to thinning due to direct forcing. Since melt pond coverage is one of the key parameters controlling the albedo of sea ice, understanding the mechanisms that control the distribution of pond coverage will help improve large-scale model parameterizations and sea ice forecasts in a warming climate.

  2. Modeling and simulation of melt-layer erosion during a plasma disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Belan, V.; Konkashbaev, I.; Nikandrov, L.; Safronov, V.; Zhitlukhin, A.; Litunovsky, V.

    1997-01-01

    Metallic plasma-facing components (PFCs) e.g. beryllium and tungsten, will be subjected to severe melting during plasma instabilities such as disruptions, edge-localized modes and high power excursions. Because of the greater thickness of the resulting melt layers relative to that of the surface vaporization, the potential loss of the developing melt-layer can significantly shorten PFC lifetime, severely contaminate the plasma and potentially prevent successful operation of the tokamak reactor. Mechanisms responsible for melt-layer loss during plasma instabilities are being modeled and evaluated. Of particular importance are hydrodynamic instabilities developed in the liquid layer due to various forces such as those from magnetic fields, plasma impact momentum, vapor recoil and surface tension. Another mechanism found to contribute to melt-layer splashing loss is volume bubble boiling, which can result from overheating of the liquid layer. To benchmark these models, several new experiments were designed and performed in different laboratory devices for this work; the SPLASH codes) are generally in good agreement with the experimental results. The effect of in-reactor disruption conditions, which do not exist in simulation experiments, on melt-layer erosion is discussed. (orig.)

  3. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    International Nuclear Information System (INIS)

    Thakre, S.; Ma, W.

    2013-08-01

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  4. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  5. A rheological model for glassforming silicate melts in the systems CAS, MAS, MCAS

    International Nuclear Information System (INIS)

    Giordano, Daniele; Russell, J K

    2007-01-01

    Viscosity is the single most important property governing the efficacy, rates, and nature of melt transport. Viscosity is intimately related to the structure and thermodynamics properties of the melts and is a reflection of the mechanisms of single atoms slipping over potential energy barriers. The ability to predict melt viscosity accurately is, therefore, of critical importance for gaining new insights into the structure of silicate melts. Simple composition melts, having a reduced number of components, offer an advantage for understanding the relationships between the chemical composition, structural organization and the rheological properties of a melt. Here we have compiled a large database of ∼970 experimental measurements of melt viscosity for the simple chemical systems MAS, CAS and MCAS. These data are used to create a single chemical model for predicting the non-Arrhenian viscosity as a function of temperature (T) and composition (X) across the entire MCAS system. The T-dependence of viscosity is accounted for by the three parameters in each of the model functions: (i) Vogel-Fulcher-Tamman (VFT); (ii) Adam-Gibbs (AG); and (iii) Avramov (AV). The literature shows that, in these systems, viscosity converges to a common value of the pre-exponential factors (A) that can be assumed to be independent of composition. The other two adjustable parameters in each equation are expanded to capture the effects of composition. The resulting models are continuous across T-X space. The values and implications of the optimal parameters returned for each model are compared and discussed. A similar approach is likely to be applicable to a variety of non-silicate multicomponent glassforming systems

  6. Size and temperature consideration in the liquid layer growth from nanovoids and the melting model construction

    International Nuclear Information System (INIS)

    Li, H.; Liang, X.H.; Li, M.

    2014-01-01

    A new model for the solid melting point T m (D) from nanovoids is proposed through considering the liquid layer growth behavior. This model, which does not have any adjustable parameter, introduces the classical thermodynamic treatment, i.e., the liquid nucleation and growth theory, for nanoparticle melting. With increased void diameter D, T m (D) approaches to T m0 . Moreover, T m (D) > T m0 for a small void (T m0 is the bulk melting point). In other words, the solid can be significantly superheated especially when D decreases, even if the difference of interface energy is larger than zero. This finding can be expected from the negatively curved surface of the void. The model predictions are consistent with the molecular dynamic (MD) simulation results for argon solids. Moreover, the growth of liquid layer from void surface relies on both size and temperature, which directly determine liquid layer thickness, and only when liquid layer thickness reaches to a critical value, can void become instable. - Highlights: • A united model for the crystal melting point from nanovoids is established. • Melting point increases with decreased void size. • The result is expected from the negatively curved surface of the void. • The prediction is agreed well with the MD simulation results

  7. Seismic risk assessment of a BWR: status report

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

    1985-02-01

    The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

  8. Geotechnical modeling of high-level nuclear waste disposal by rock melting

    International Nuclear Information System (INIS)

    Heuze, F.E.

    1981-12-01

    A new strategy has been developed for the geotechnical modeling of nuclear waste disposal by rock melting (DRM). Three seeparate tasks were performed to reach this objective: a review of the four scenarios which have been proposed for DRM, to date; an evaluation of computer-based numerical models which could be used to analyze the mechanical, thermal, and hydraulic processes involved in DRM; and a critical review of rock mass properties which are relevant to the design and safety of waste disposal by rock melting. It is concluded that several geotechnical aspects of DRM can be studied realistically with current state-of-the-art model capabilities and knowledge of material properties. The next step in the feasibility study of DRM should be a best-estimate calculation of the four cavity-melt and canister-burial concepts. These new analyses will indicate the most critical areas for subsequent research

  9. The use of the partial coherence function technique for the investigation of BWR noise dynamics

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1983-01-01

    The extensive experimental investigations, at the last time, indicate that the partial coherence function technique can be a powerful method of the investigation of BWR noise dynamics. Symple BWR noise dynamics model for the global noise study, based on different noise phenomena, is proposed in this paper. (author)

  10. Level-Ice Melt Ponds in the Los Alamos Sea Ice Model, CICE

    Science.gov (United States)

    2012-12-06

    terms obtained using the Bitz and Lips- comb (1999) thermodynamic model. The thickness distribution ( Thorndike et al., 1975) employs 5 ice thickness...D.L., 2004. A model of melt pond evolution on sea ice. J. Geophys. Res. 109, C12007. http://dx.doi.org/10.1029/2004JC002361. Thorndike , A.S., Rothrock

  11. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant; Implementacion digital, simulacion y pruebas en MATLAB de los modelos de la linea de vapor, las turbinas y el regulador de presion de una central Nucleoelectrica tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)

    2004-07-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  12. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator

    International Nuclear Information System (INIS)

    Sanchez S, R.A.

    2003-01-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  13. Effect of nonlinear void reactivity on bifurcation characteristics of a lumped-parameter model of a BWR: A study relevant to RBMK

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2017-04-15

    Highlights: • A simplified model with nonlinear void reactivity feedback is studied. • Method of multiple scales for nonlinear analysis and oscillation characteristics. • Second order void reactivity dominates in determining system dynamics. • Opposing signs of linear and quadratic void reactivity enhances global safety. - Abstract: In the present work, the effect of nonlinear void reactivity on the dynamics of a simplified lumped-parameter model for a boiling water reactor (BWR) is investigated. A mathematical model of five differential equations comprising of neutronics and thermal-hydraulics encompassing the nonlinearities associated with both the reactivity feedbacks and the heat transfer process has been used. To this end, we have considered parameters relevant to RBMK for which the void reactivity is known to be nonlinear. A nonlinear analysis of the model exploiting the method of multiple time scales (MMTS) predicts the occurrence of the two types of Hopf bifurcation, namely subcritical and supercritical, leading to the evolution of limit cycles for a range of parameters. Numerical simulations have been performed to verify the analytical results obtained by MMTS. The study shows that the nonlinear reactivity has a significant influence on the system dynamics. A parametric study with varying nominal reactor power and operating conditions in coolant channel has also been performed which shows the effect of change in concerned parameter on the boundary between regions of sub- and super-critical Hopf bifurcations in the space constituted by the two coefficients of reactivities viz. the void and the Doppler coefficient of reactivities. In particular, we find that introduction of a negative quadratic term in the void reactivity feedback significantly increases the supercritical region and dominates in determining the system dynamics.

  14. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  15. Modelling and parameterizing the influence of tides on ice-shelf melt rates

    Science.gov (United States)

    Jourdain, N.; Molines, J. M.; Le Sommer, J.; Mathiot, P.; de Lavergne, C.; Gurvan, M.; Durand, G.

    2017-12-01

    Significant Antarctic ice sheet thinning is observed in several sectors of Antarctica, in particular in the Amundsen Sea sector, where warm circumpolar deep waters affect basal melting. The later has the potential to trigger marine ice sheet instabilities, with an associated potential for rapid sea level rise. It is therefore crucial to simulate and understand the processes associated with ice-shelf melt rates. In particular, the absence of tides representation in ocean models remains a caveat of numerous ocean hindcasts and climate projections. In the Amundsen Sea, tides are relatively weak and the melt-induced circulation is stronger than the tidal circulation. Using a regional 1/12° ocean model of the Amundsen Sea, we nonetheless find that tides can increase melt rates by up to 36% in some ice-shelf cavities. Among the processes that can possibly affect melt rates, the most important is an increased exchange at the ice/ocean interface resulting from the presence of strong tidal currents along the ice drafts. Approximately a third of this effect is compensated by a decrease in thermal forcing along the ice draft, which is related to an enhanced vertical mixing in the ocean interior in presence of tides. Parameterizing the effect of tides is an alternative to the representation of explicit tides in an ocean model, and has the advantage not to require any filtering of ocean model outputs. We therefore explore different ways to parameterize the effects of tides on ice shelf melt. First, we compare several methods to impose tidal velocities along the ice draft. We show that getting a realistic spatial distribution of tidal velocities in important, and can be deduced from the barotropic velocities of a tide model. Then, we explore several aspects of parameterized tidal mixing to reproduce the tide-induced decrease in thermal forcing along the ice drafts.

  16. Photoinduced charge-order melting dynamics in a one-dimensional interacting Holstein model

    Science.gov (United States)

    Hashimoto, Hiroshi; Ishihara, Sumio

    2017-07-01

    Transient quantum dynamics in an interacting fermion-phonon system are investigated with a focus on a charge order (CO) melting after a short optical-pulse irradiation and the roles of the quantum phonons in the transient dynamics. A spinless-fermion model in a one-dimensional chain coupled with local phonons is analyzed numerically. The infinite time-evolving block decimation algorithm is adopted as a reliable numerical method for one-dimensional quantum many-body systems. Numerical results for the photoinduced CO melting dynamics without phonons are well interpreted by the soliton picture for the CO domains. This interpretation is confirmed by numerical simulation of an artificial local excitation and the classical soliton model. In the case of large phonon frequencies corresponding to the antiadiabatic condition, CO melting is induced by propagations of the polaronic solitons with the renormalized soliton velocity. On the other hand, in the case of small phonon frequencies corresponding to the adiabatic condition, the first stage of the CO melting dynamics occurs due to the energy transfer from the fermionic to phononic systems, and the second stage is brought about by the soliton motions around the bottom of the soliton band. The analyses provide a standard reference for photoinduced CO melting dynamics in one-dimensional many-body quantum systems.

  17. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  18. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

    Post, J.S.; Karner, E.F.; Stratman, R.A.

    1984-01-01

    This chapter describes plans for dealing with reactor accidents developed by the Boiling Water Reactor (BWR) Owners' Group in response to post-Three Mile Island US NRC requirements. The devised Emergency Procedure Guidelines (EPGs), applicable to all BWRs, are symptom-based rather than event-based. According to the EPGs, the operator does not need to identify what event is occurring in the plant in order to decide what action to take, but need only observe the symptoms (values and trends of key control parameters) which exist and take appropriate action to control these symptoms. The original objective was to provide reactor operator guidance in responding to a small break loss-of-coolant accident (LOCA), but subsequent revisions have included other types of reactor accidents. Topics considered include the reactor pressure vessel (RPV) control guideline, the primary containment control guideline, the secondary containment control guideline, the radioactivity release control guideline, multiple failures vs. the design basis, safe limits vs. technical specifications, the technical status, licensing, and implementation. The EPGs are based upon maintaining both adequate core cooling and primary containment integrity

  19. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  20. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  1. Outlooks for mathematical modelling of the glass melting process

    Energy Technology Data Exchange (ETDEWEB)

    Waal, H. de [TNO Institute of Applied Physics, Delft (Netherlands)

    1997-12-31

    Mathematical modelling is nowadays a standard tool for major producers of float glass, T.V. glass and fiberglass. Also for container glass furnaces, glass tank modelling proves to be a valuable method to optimize process conditions. Mathematical modelling is no longer just a way to visualize the flow patterns and to provide data on heat transfer. It can also predict glass quality in relation to process parameters, because all chemical and physical phenomena are included in the latest generation of models, based on experimental and theoretical research on these phenomena.

  2. Development and assessment of modular models of calculation for the interpretation of rod-melting experiments

    International Nuclear Information System (INIS)

    Tuerk, W.

    1980-01-01

    By the example of recalculations of rod-melting experiment it is shown how a modular simulation model for complex systems can be formulated within the scope of RSYST1. The procedure of code development as well as the physical and numerical methods and approximations of the simulation model are described. To each important physical process a code module is assigned. The individual moduls describe heat production, rod heat-up, rod oxidation, rod environment, rod deformation by thermal expansion and can buckling, melting of the rod, rod failure, and flowing off of the melted mass. A comparison of the results for the overall model with the result of different experiments indicates that the phenomena during heat-up and melting of the rod are treated in agreement with the experiments. The results of the calculation model and its submodels are thus largely supported by experiments. Therefore further predictions with a high level of confidence can be made with the model within the scope of reactor safety research. (orig.) [de

  3. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

  4. Models for mean bonding length, melting point and lattice thermal expansion of nanoparticle materials

    Energy Technology Data Exchange (ETDEWEB)

    Omar, M.S., E-mail: dr_m_s_omar@yahoo.com [Department of Physics, College of Science, University of Salahaddin-Erbil, Arbil, Kurdistan (Iraq)

    2012-11-15

    Graphical abstract: Three models are derived to explain the nanoparticles size dependence of mean bonding length, melting temperature and lattice thermal expansion applied on Sn, Si and Au. The following figures are shown as an example for Sn nanoparticles indicates hilly applicable models for nanoparticles radius larger than 3 nm. Highlights: ► A model for a size dependent mean bonding length is derived. ► The size dependent melting point of nanoparticles is modified. ► The bulk model for lattice thermal expansion is successfully used on nanoparticles. -- Abstract: A model, based on the ratio number of surface atoms to that of its internal, is derived to calculate the size dependence of lattice volume of nanoscaled materials. The model is applied to Si, Sn and Au nanoparticles. For Si, that the lattice volume is increases from 20 Å{sup 3} for bulk to 57 Å{sup 3} for a 2 nm size nanocrystals. A model, for calculating melting point of nanoscaled materials, is modified by considering the effect of lattice volume. A good approach of calculating size-dependent melting point begins from the bulk state down to about 2 nm diameter nanoparticle. Both values of lattice volume and melting point obtained for nanosized materials are used to calculate lattice thermal expansion by using a formula applicable for tetrahedral semiconductors. Results for Si, change from 3.7 × 10{sup −6} K{sup −1} for a bulk crystal down to a minimum value of 0.1 × 10{sup −6} K{sup −1} for a 6 nm diameter nanoparticle.

  5. Models for mean bonding length, melting point and lattice thermal expansion of nanoparticle materials

    International Nuclear Information System (INIS)

    Omar, M.S.

    2012-01-01

    Graphical abstract: Three models are derived to explain the nanoparticles size dependence of mean bonding length, melting temperature and lattice thermal expansion applied on Sn, Si and Au. The following figures are shown as an example for Sn nanoparticles indicates hilly applicable models for nanoparticles radius larger than 3 nm. Highlights: ► A model for a size dependent mean bonding length is derived. ► The size dependent melting point of nanoparticles is modified. ► The bulk model for lattice thermal expansion is successfully used on nanoparticles. -- Abstract: A model, based on the ratio number of surface atoms to that of its internal, is derived to calculate the size dependence of lattice volume of nanoscaled materials. The model is applied to Si, Sn and Au nanoparticles. For Si, that the lattice volume is increases from 20 Å 3 for bulk to 57 Å 3 for a 2 nm size nanocrystals. A model, for calculating melting point of nanoscaled materials, is modified by considering the effect of lattice volume. A good approach of calculating size-dependent melting point begins from the bulk state down to about 2 nm diameter nanoparticle. Both values of lattice volume and melting point obtained for nanosized materials are used to calculate lattice thermal expansion by using a formula applicable for tetrahedral semiconductors. Results for Si, change from 3.7 × 10 −6 K −1 for a bulk crystal down to a minimum value of 0.1 × 10 −6 K −1 for a 6 nm diameter nanoparticle.

  6. SCDAP/RELAP5 modeling of movement of melted material through porous debris in lower head

    International Nuclear Information System (INIS)

    Siefken, L. J.; Harvego, E. A.

    2000-01-01

    A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has been implemented into the SCDAP/RELAP5 code for the analysis of the integrity of LWR lower heads during severe accidents. The results of the model indicate that melted core plate material may permeate to near the bottom of a 1m deep hot porous debris bed supported by the lower head. The presence of the relocated core plate material was calculated to cause a 12% increase in the heat flux on the external surface of the lower head

  7. The synergy of modeling and novel experiments for melt crystal growth research

    Science.gov (United States)

    Derby, Jeffrey J.

    2018-05-01

    Computational modeling and novel experiments, when performed together, can enable the identification of new, fundamental mechanisms important for the growth of bulk crystals from the melt. In this paper, we present a compelling example of this synergy via the discovery of previously unascertained physical mechanisms that govern the engulfment of silicon carbide particles during the growth of crystalline silicon.

  8. Melting behavior of a model molecular crystalline GeI4

    International Nuclear Information System (INIS)

    Fuchizaki, Kazuhiro; Asano, Yuta

    2015-01-01

    A model molecular crystalline GeI 4 was examined using molecular dynamics simulation. The model was constructed in such a way that rigid tetrahedral molecules interact with each other via Lennard-Jones potentials whose centers are located at the vertices of a tetrahedron. Because no other interaction that can “soften” the intermolecular interaction was introduced, the melting curve of the model crystalline material does not exhibit the anomaly that was found for the real substance. However, the current investigation is useful in that it could settle the upper bound of pressure below which the model can predict properties of the molecular liquid. Moreover, singularity-free nature of the melting curve allowed us to analytically treat the melting curve in the light of the Kumari-Dass-Kechin equation. As a result, we could definitely conclude that the well-known Simon equation for the melting curve is merely an approximate expression. The condition for the validity of Simon’s equation was identified. (author)

  9. BWR Services maintenance training program

    International Nuclear Information System (INIS)

    Cox, J.H.; Chittenden, W.F.

    1979-01-01

    BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

  10. An analysis of molten-corium-induced failure of drain pipes in BWR Mark 2 containments

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.; Podowski, M.Z.

    1991-01-01

    This study has focused on mechanistic simulation and analysis of potential failure modes for inpedestal drywell drain pipes in the Limerick boiling water reactor (BWR) Mark 2 containment. Physical phenomena related to surface tension breakdown, heatup, melting, ablation, crust formation and failure, and core material relocation into drain pipes with simultaneous melting of pipe walls were modeled and analyzed. The results of analysis have been used to assess the possibility of drain pipe failure and the resultant loss of pressure-suppression capability. Estimates have been made for the timing and amount of molten corium released to the wetwell. The study has revealed that significantly different melt progression sequences can result depending upon the failure characteristics of the frozen metallic crust which forms over the drain cover during the initial stages of debris pour. Another important result is that it can take several days for the molten fuel to ablate the frozen metallic debris layer -- if the frozen layer has cooled below 1100 K before fuel attack. 10 refs., 3 figs., 4 tabs

  11. What Models and Satellites Tell Us (and Don't Tell Us) About Arctic Sea Ice Melt Season Length

    Science.gov (United States)

    Ahlert, A.; Jahn, A.

    2017-12-01

    Melt season length—the difference between the sea ice melt onset date and the sea ice freeze onset date—plays an important role in the radiation balance of the Arctic and the predictability of the sea ice cover. However, there are multiple possible definitions for sea ice melt and freeze onset in climate models, and none of them exactly correspond to the remote sensing definition. Using the CESM Large Ensemble model simulations, we show how this mismatch between model and remote sensing definitions of melt and freeze onset limits the utility of melt season remote sensing data for bias detection in models. It also opens up new questions about the precise physical meaning of the melt season remote sensing data. Despite these challenges, we find that the increase in melt season length in the CESM is not as large as that derived from remote sensing data, even when we account for internal variability and different definitions. At the same time, we find that the CESM ensemble members that have the largest trend in sea ice extent over the period 1979-2014 also have the largest melt season trend, driven primarily by the trend towards later freeze onsets. This might be an indication that an underestimation of the melt season length trend is one factor contributing to the generally underestimated sea ice loss within the CESM, and potentially climate models in general.

  12. Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

    International Nuclear Information System (INIS)

    Hyman, C.R.

    1988-01-01

    Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,

  13. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  14. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  15. A Thermal Hydraulic Model of Melt Lubrication in Railgun Armatures

    National Research Council Canada - National Science Library

    Kothmann, R

    2003-01-01

    ... wear of 7075 aluminum sliding against ETP copper for face pressures ranging from 6 to 22 ksi. Discrepancies between calculated and experimental results are attributed to uncertainties in modeling the complex phase change behavior of aluminum alloy 7075 and uncertain conditions at the rail interface.

  16. A vapour bubble collapse model to describe the fragmentation of low-melting materials

    International Nuclear Information System (INIS)

    Benz, R.; Schober, P.

    1977-11-01

    By means of a model, the fragmentation of a hot melt of metal in consequence of collapsing vapour-bubbles is investigated. In particular the paper deals with the development of the physical model-ideas for calculation of the temperature of contact that adjusts between the temperature of the melt and the coolant, of the waiting-time until bubble-nucleation occurs and of the maximal obtainable vapour-bubble-radius in dependence of the coolant-temperature. After that follows the description of the computing-program belonging to this model and of the results of an extensive parameter-study. The study examined the influence of the temperature of melt and coolant, the melted mass, the nucleation-site-density, the average maximum bubble-radius, the duration of film-breakdown and the coefficient of heat-transition. The calculation of the process of fragmentation turns out to be according to expectation, whereas the duration of this process seems to be somewhat too long. The dependence of the surface-enlargement on the subcooling of the water-bath and the initial temperature of the melt is not yet reproduced satisfactorily by the model. The reasons for this are the temperature-increase of the water-bath as well as the fact that the coupling of heat-flux-density and nucleation-site-density are not taken into consideration. Further improvement of the model is necessary and may improve the results in the sense of the experimental observations. (orig.) [de

  17. The BWR owners' group planning guide for life extension

    International Nuclear Information System (INIS)

    Smith, S.K.; Lehnert, D.F.; Locke, R.K.

    1991-01-01

    Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

  18. A finite volume alternate direction implicit approach to modeling selective laser melting

    DEFF Research Database (Denmark)

    Hattel, Jesper Henri; Mohanty, Sankhya

    2013-01-01

    Over the last decade, several studies have attempted to develop thermal models for analyzing the selective laser melting process with a vision to predict thermal stresses, microstructures and resulting mechanical properties of manufactured products. While a holistic model addressing all involved...... to accurately simulate the process, are constrained by either the size or scale of the model domain. A second challenging aspect involves the inclusion of non-linear material behavior into the 3D implicit FE models. An alternating direction implicit (ADI) method based on a finite volume (FV) formulation...... is proposed for modeling single-layer and few-layers selective laser melting processes. The ADI technique is implemented and applied for two cases involving constant material properties and non-linear material behavior. The ADI FV method consume less time while having comparable accuracy with respect to 3D...

  19. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

    2007-01-01

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  20. The possibility and the effects of a steam explosion in the BWR lower head on recriticality of a BWR core

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.

    2002-12-01

    The report describes an analysis considering a BWR postulated severe accident scenario during which the late vessel automatic depressurization brings the water below the level of the bottom core plate. The subsequent lack of ECCS leads to core heat up during which the control rods melt and the melt deposits on the core plate. At that point of time in the scenario, the core fuel bundles are still intact and the Zircaloy clad oxidation is about to start. The objective of the study is to provide the conditions of reflood into the hot core due to the level swell or a slug delivered from the lower head as the control rod melt drops into the water. These conditions are employed in the neutronic analysis with the RECRIT code to determine if the core recriticality may be achieved. (au)

  1. Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR

    Energy Technology Data Exchange (ETDEWEB)

    Galushin, Sergey, E-mail: galushin@kth.se; Kudinov, Pavel, E-mail: pkudinov@kth.se

    2016-12-15

    Highlights: • A data base of the debris properties in lower plenum generated using MELCOR code. • The timing of safety systems has significant effect on the relocated debris properties. • Loose coupling between core relocation and vessel failure analyses was established. - Abstract: Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario

  2. Analysis of ex-vessel melt jet breakup and coolability. Part 1: Sensitivity on model parameters and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kiyofumi; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Hwang, Byoungcheol; Jung, Woo Hyun

    2016-06-15

    Highlights: • Application of JASMINE code to melt jet breakup and coolability in APR1400 condition. • Coolability indexes for quasi steady state breakup and cooling process. • Typical case in complete breakup/solidification, film boiling quench not reached. • Significant impact of water depth and melt jet size; weak impact of model parameters. - Abstract: The breakup of a melt jet falling in a water pool and the coolability of the melt particles produced by such jet breakup are important phenomena in terms of the mitigation of severe accident consequences in light water reactors, because the molten and relocated core material is the primary heat source that governs the accident progression. We applied a modified version of the fuel–coolant interaction simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA) to a plant scale simulation of melt jet breakup and cooling assuming an ex-vessel condition in the APR1400, a Korean advanced pressurized water reactor. Also, we examined the sensitivity on seven model parameters and five initial/boundary condition variables. The results showed that the melt cooling performance of a 6 m deep water pool in the reactor cavity is enough for removing the initial melt enthalpy for solidification, for a melt jet of 0.2 m initial diameter. The impacts of the model parameters were relatively weak and that of some of the initial/boundary condition variables, namely the water depth and melt jet diameter, were very strong. The present model indicated that a significant fraction of the melt jet is not broken up and forms a continuous melt pool on the containment floor in cases with a large melt jet diameter, 0.5 m, or a shallow water pool depth, ≤3 m.

  3. Advanced One-Dimensional Entrained-Flow Gasifier Model Considering Melting Phenomenon of Ash

    Directory of Open Access Journals (Sweden)

    Jinsu Kim

    2018-04-01

    Full Text Available A one-dimensional model is developed to represent the ash-melting phenomenon, which was not considered in the previous one-dimensional (1-D entrained-flow gasifier model. We include sensible heat of slag and the fusion heat of ash in the heat balance equation. To consider the melting of ash, we propose an algorithm that calculates the energy balance for three scenarios based on temperature. We also use the composition and the thermal properties of anorthite mineral to express ash. gPROMS for differential equations is used to solve this algorithm in a simulation; the results include coal conversion, gas composition, and temperature profile. Based on the Texaco pilot plant gasifier, we validate our model. Our results show good agreement with previous experimental data. We conclude that the sensible heat of slag and the fusion heat of ash must be included in the entrained flow gasifier model.

  4. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    International Nuclear Information System (INIS)

    Tran, Chi Thanh

    2009-09-01

    indispensable for scrutinizing flow physics, on the other hand, the validated CFD method can be used to generate necessary data for validation of the accident analysis models. Given the insights gained from the CFD study, physics-based models and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a LWR during the late phase of an in-vessel core melt progression. To describe natural convection heat transfer in an internally heated volume, and molten metal layer heated from below and cooled from the top (and side) walls, the Effective Convectivity Models (ECM) are developed and implemented in a commercial CFD code. The ECM uses directional heat transfer characteristic velocities to transport the heat to cooled boundaries. The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables 3D heat transfer simulations of a homogeneous (and stratified) melt pool formed in the LWR lower head. In order to describe phase-change heat transfer associated with core debris or binary mixture (e.g. in a molten metal layer), a temperature-based enthalpy formulation is employed in the Phase-change ECM (so called the PECM). The PECM is capable to represent natural convection heat transfer in a mushy zone. Simple formulation of the PECM method allows implementing different models of mushy zone heat transfer for non-eutectic mixtures. For a non-eutectic binary mixture, compositional convection associated with concentration gradients can be taken into account. The developed models are validated against both existing experimental data and the CFD-generated data. ECM and PECM simulations show a superior computational efficiency compared to the CFD simulation method. The ECM and PECM methods are applied to predict thermal loads imposed on the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in a Boiling Water Reactor (BWR

  5. Modeling of beam-target interaction during pulsed electron beam ablation of graphite: Case of melting

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Muddassir, E-mail: mx1_ali@laurentian.ca; Henda, Redhouane

    2017-02-28

    Highlights: • Modeling of ablation stage induced during pulsed electron beam ablation (PEBA). • Thermal model to describe heating, melting and vaporization of a graphite target. • Model results show good accordance with reported data in the literature. - Abstract: A one-dimensional thermal model based on a two-stage heat conduction equation is employed to investigate the ablation of graphite target during nanosecond pulsed electron beam ablation. This comprehensive model accounts for the complex physical phenomena comprised of target heating, melting and vaporization upon irradiation with a polyenergetic electron beam. Melting and vaporization effects induced during ablation are taken into account by introducing moving phase boundaries. Phase transition induced during ablation is considered through the temperature dependent thermodynamic properties of graphite. The effect of electron beam efficiency, power density, and accelerating voltage on ablation is analyzed. For an electron beam operating at an accelerating voltage of 15 kV and efficiency of 0.6, the model findings show that the target surface temperature can reach up to 7500 K at the end of the pulse. The surface begins to melt within 25 ns from the pulse start. For the same process conditions, the estimated ablation depth and ablated mass per unit area are about 0.60 μm and 1.05 μg/mm{sup 2}, respectively. Model results indicate that ablation takes place primarily in the regime of normal vaporization from the surface. The results obtained at an accelerating voltage of 15 kV and efficiency factor of 0.6 are satisfactorily in good accordance with available experimental data in the literature.

  6. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  7. Proposed model for fuel-coolant mixing during a core-melt accident

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1983-01-01

    If complete failure of normal and emergency coolant flow occurs in a light water reactor, fission product decay heat would eventually cause melting of the reactor fuel and cladding. The core melt may then slump into the lower plenum and later into the reactor cavity and contact residual liquid water. A model is proposed to describe the fuel-coolant mixing process upon contact. The model is compared to intermediate scale experiments being conducted at Sandia. The modelling of this mixing process will aid in understanding three important processes: (1) fuel debris sizes upon quenching in water, (2) the hydrogen source term during fuel quench, and (3) the rate of steam production. Additional observations of Sandia data indicate that the steam explosion is affected by this mixing process

  8. Effect Of Turbulence Modelling In Numerical Analysis Of Melting Process In An Induction Furnace

    Directory of Open Access Journals (Sweden)

    Buliński P.

    2015-09-01

    Full Text Available In this paper, the velocity field and turbulence effects that occur inside a crucible of a typical induction furnace were investigated. In the first part of this work, a free surface shape of the liquid metal was measured in a ceramic crucible. Then a numerical model of aluminium melting process was developed. It took into account coupling of electromagnetic and thermofluid fields that was performed using commercial codes. In the next step, the sensitivity analysis of turbulence modelling in the liquid domain was performed. The obtained numerical results were compared with the measurement data. The performed analysis can be treated as a preliminary approach for more complex mathematical modelling for the melting process optimisation in crucible induction furnaces of different types.

  9. Modeling radar backscattering from melting snowflakes using spheroids with nonuniform distribution of water

    International Nuclear Information System (INIS)

    Tyynelä, Jani; Leinonen, Jussi; Moisseev, Dmitri; Nousiainen, Timo; Lerber, Annakaisa von

    2014-01-01

    In a number of studies it is reported that at the early stages, melting of aggregate snowflakes is enhanced at lower parts. In this paper, the manifestation of the resulting nonuniform distribution of water is studied for radar backscattering cross sections at C, Ku, Ka and W bands. The melting particles are described as spheroids with a mixture of water and air at the bottom part of the particle and a mixture of ice and air at the upper part. The radar backscattering is modeled using the discrete-dipole approximation in a horizontally pointing geometry. The results are compared to the T-matrix method, Mie theory, and the Rayleigh approximation using the Maxwell Garnett mixing formula. We find that the differential reflectivity and the linear depolarization ratio show systematic differences between the discrete-dipole approximation and the T-matrix method, but that the differences are relatively small. The horizontal cross sections show only small differences between the methods with the aspect ratio and the presence of resonance peaks having a larger effect on it than the nonuniform distribution of water. Overall, the effect of anisotropic distribution of water, reported for early stages of melting, is not significant for radar observations at the studied frequencies. -- Highlights: • We model backscattering from spheroidal melting snowflakes at C, Ku, Ka, and W bands. • We study the effect of anisotropic distribution of meltwater in the snow particles. • We find systematic, but relatively small differences for the backscattering properties. • We find that the aspect ratio and resonance peaks have a bigger effect than anisotropic distribution of water. • Anisotropic distribution of water is not significant for radar observations at early stages of melting

  10. Modeling and Experimental Validation of the Electron Beam Selective Melting Process

    Directory of Open Access Journals (Sweden)

    Wentao Yan

    2017-10-01

    Full Text Available Electron beam selective melting (EBSM is a promising additive manufacturing (AM technology. The EBSM process consists of three major procedures: ① spreading a powder layer, ② preheating to slightly sinter the powder, and ③ selectively melting the powder bed. The highly transient multi-physics phenomena involved in these procedures pose a significant challenge for in situ experimental observation and measurement. To advance the understanding of the physical mechanisms in each procedure, we leverage high-fidelity modeling and post-process experiments. The models resemble the actual fabrication procedures, including ① a powder-spreading model using the discrete element method (DEM, ② a phase field (PF model of powder sintering (solid-state sintering, and ③ a powder-melting (liquid-state sintering model using the finite volume method (FVM. Comprehensive insights into all the major procedures are provided, which have rarely been reported. Preliminary simulation results (including powder particle packing within the powder bed, sintering neck formation between particles, and single-track defects agree qualitatively with experiments, demonstrating the ability to understand the mechanisms and to guide the design and optimization of the experimental setup and manufacturing process.

  11. Production, pathways and budgets of melts in mid-ocean ridges: An enthalpy based thermo-mechanical model

    Science.gov (United States)

    Mandal, Nibir; Sarkar, Shamik; Baruah, Amiya; Dutta, Urmi

    2018-04-01

    Using an enthalpy based thermo-mechanical model we provide a theoretical evaluation of melt production beneath mid-ocean ridges (MORs), and demonstrate how the melts subsequently develop their pathways to sustain the major ridge processes. Our model employs a Darcy idealization of the two-phase (solid-melt) system, accounting enthalpy (ΔH) as a function of temperature dependent liquid fraction (ϕ). Random thermal perturbations imposed in this model set in local convection that drive melts to flow through porosity controlled pathways with a typical mushroom-like 3D structure. We present across- and along-MOR axis model profiles to show the mode of occurrence of melt-rich zones within mushy regions, connected to deeper sources by single or multiple feeders. The upwelling of melts experiences two synchronous processes: 1) solidification-accretion, and 2) eruption, retaining a large melt fraction in the framework of mantle dynamics. Using a bifurcation analysis we determine the threshold condition for melt eruption, and estimate the potential volumes of eruptible melts (∼3.7 × 106 m3/yr) and sub-crustal solidified masses (∼1-8.8 × 106 m3/yr) on an axis length of 500 km. The solidification process far dominates over the eruption process in the initial phase, but declines rapidly on a time scale (t) of 1 Myr. Consequently, the eruption rate takes over the solidification rate, but attains nearly a steady value as t > 1.5 Myr. We finally present a melt budget, where a maximum of ∼5% of the total upwelling melt volume is available for eruption, whereas ∼19% for deeper level solidification; the rest continue to participate in the sub-crustal processes.

  12. Performances of the snow accumulation melting model SAMM: results in the Northern Apennines test area

    Science.gov (United States)

    Lagomarsino, Daniela; Martelloni, Gianluca; Segoni, Samuele; Catani, Filippo; Fanti, Riccardo

    2013-04-01

    In this work we propose a snow accumulation-melting model (SAMM) to forecast the snowpack height and we compare the results with a simple temperature index model and an improved version of the latter.For this purpose we used rainfall, temperature and snowpack thickness 5-years data series from 7 weather stations in the Northern Apennines (Emilia Romagna Region, Italy). SAMM is based on two modules modelling the snow accumulation and the snowmelt processes. Each module is composed by two equations: a mass conservation equation is solved to model snowpack thickness and an empirical equation is used for the snow density. The processes linked to the accumulation/depletion of the snowpack (e.g. compression of the snowpack due to newly fallen snow and effects of rainfall) are modelled identifying limiting and inhibitory factors according to a kinetic approach. The model depends on 13 empirical parameters, whose optimal values were defined with an optimization algorithm (simplex flexible) using calibration measures of snowpack thickness. From an operational point of view, SAMM uses as input data only temperature and rainfall measurements, bringing the additional advantage of a relatively easy implementation. In order to verify the improvement of SAMM with respect to a temperature-index model, the latter was applied considering, for the amount of snow melt, the following equation: M = fm(T-T0), where M is hourly melt, fm is the melting factor and T0 is a threshold temperature. In this case the calculation of the depth of the snowpack requires the use of 3 parameters: fm, T0 and ?0 (the mean density of the snowpack). We also performed a simulation by replacing the SAMM melting module with the above equation and leaving unchanged the accumulation module: in this way we obtained a model with 9 parameters. The simulations results suggest that any further extension of the simple temperature index model brings some improvements with a consequent decrease of the mean error

  13. Model for melt blockage (slug) relocation and physico-chemical interactions during core degradation under severe accident conditions

    International Nuclear Information System (INIS)

    Veshchunov, M.S.; Shestak, V.E.

    2008-01-01

    The model describing massive melt blockage (slug) relocation and physico-chemical interactions with steam and surrounding fuel rods of a bundle is developed on the base of the observations in the CORA tests. Mass exchange owing to slug oxidation and fuel rods dissolution is described by the previously developed 2D model for the molten pool oxidation. Heat fluxes in oxidising melt along with the oxidation heat effect at the melt relocation front are counterbalanced by the heat losses in the surrounding media and the fusion heat effect of the Zr claddings attacked by the melt. As a result, the slug relocation velocity is calculated from the heat flux matches at the melt propagation front (Stefan problem). A numerical module simulating the slug behaviour is developed by tight coupling of the heat and mass exchange modules. The new model demonstrates a reasonable capability to simulate the main features of the massive slug behaviour observed in the CORA-W1 test

  14. Hydraulic modeling and simulation of a System Division of Essential Service Water in a BWR plant with Flow master

    International Nuclear Information System (INIS)

    Vegazo Juzgado, L.; Rodriguez Garcia, G. M.; Mota Coloma, M.

    2012-01-01

    At the conclusion of the project can say that Flow master is a simulation tool that allows you to create your model from a library of components and obtain useful results from the point of view of the operation, engineering and maintenance. Compared to previous software from the point of view of use, can comment that Flow master is a tool which has an intuitive and user-friendly interaction between the user and the program thus facilitating the modeling of the system and definition of the components of same.

  15. Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

    International Nuclear Information System (INIS)

    Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

    1984-01-01

    This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

  16. Study of melting of molecular crystals by a modified Pople-Karasz model

    Science.gov (United States)

    Yazıcı, Mustafa; Özgan, Şükrü; Keskin, Mustafa

    2005-02-01

    A new modified model that combines the modified models of Chandrasekhar et al. with those of Keskin and Özgan, which are based on the Pople-Karasz theory, is applied to study the thermodynamics of melting and solid-solid transitions of molecular crystals. The thermodynamic properties of the disordered system are evaluated relative to those of the perfectly ordered one using the lowest approximation of the cluster-variation method, which is identical to the mean-field approximation. A good agreement is found between the present modified theory and the available experimental data. For melting transitions the agreement is excellent and much better than with the calculations of the Pople-Karasz theory and its previous modified theories. Approximate agreement is obtained for the solid-solid transitions. However, for these transition the experimental agreement with the present modified theory is still better than previous modified theories except at zero and low pressures.

  17. BWR modeling capability and Scale/Triton lattice-to-core integration of the Nestle nodal simulator - 331

    International Nuclear Information System (INIS)

    Galloway, J.; Hernandez, H.; Maldonado, G.I.; Jessee, M.; Popov, E.; Clarno, K.

    2010-01-01

    This article reports the status of recent and substantial enhancements made to the NESTLE nodal core simulator, a code originally developed at North Carolina State University (NCSU) of which version 5.2.1 has been available for several years through the Oak Ridge National Laboratory (ORNL) Radiation Safety Information Computational Center (RSICC) software repository. In its released and available form, NESTLE is a seasoned, well-developed and extensively tested code system particularly useful to model PWRs. In collaboration with NCSU, University of Tennessee (UT) and ORNL researchers have recently developed new enhancements for the NESTLE code, including the implementation of a two-phase drift-flux thermal hydraulic and flow redistribution model to facilitate modeling of Boiling Water Reactors (BWRs) as well as the development of an integrated coupling of SCALE/TRITON lattice physics to NESTLE so to produce an end-to-end capability for reactor simulations. These latest advancements implemented into NESTLE as well as an update of other ongoing efforts of this project are herein reported. (authors)

  18. Melting under shock compression

    International Nuclear Information System (INIS)

    Bennett, B.I.

    1980-10-01

    A simple model, using experimentally measured shock and particle velocities, is applied to the Lindemann melting formula to predict the density, temperature, and pressure at which a material will melt when shocked from room temperature and zero pressure initial conditions

  19. VOLATILECALC: A silicate melt-H2O-CO2 solution model written in Visual Basic for excel

    Science.gov (United States)

    Newman, S.; Lowenstern, J. B.

    2002-01-01

    We present solution models for the rhyolite-H2O-CO2 and basalt-H2O-CO2 systems at magmatic temperatures and pressures below ~ 5000 bar. The models are coded as macros written in Visual Basic for Applications, for use within MicrosoftR Excel (Office'98 and 2000). The series of macros, entitled VOLATILECALC, can calculate the following: (1) Saturation pressures for silicate melt of known dissolved H2O and CO2 concentrations and the corresponding equilibrium vapor composition; (2) open- and closed-system degassing paths (melt and vapor composition) for depressurizing rhyolitic and basaltic melts; (3) isobaric solubility curves for rhyolitic and basaltic melts; (4) isoplethic solubility curves (constant vapor composition) for rhyolitic and basaltic melts; (5) polybaric solubility curves for the two end members and (6) end member fugacities of H2O and CO2 vapors at magmatic temperatures. The basalt-H2O-CO2 macros in VOLATILECALC are capable of calculating melt-vapor solubility over a range of silicate-melt compositions by using the relationships provided by Dixon (American Mineralogist 82 (1997) 368). The output agrees well with the published solution models and experimental data for silicate melt-vapor systems for pressures below 5000 bar. ?? 2002 Elsevier Science Ltd. All rights reserved.

  20. Considerations in modelling the melting of fuel containing fission products and solute oxides

    International Nuclear Information System (INIS)

    Akbari, F.; Welland, M.J.; Lewis, B.J.; Thompson, W.T.

    2005-01-01

    It is well known that the oxidation of a defected fuel element by steam gives rise to an increase in O/U ratio with a consequent lowering of the incipient melting temperature. Concurrently, the hyperstoichiometry reduces the thermal conductivity thereby raising the centerline fuel pellet temperature for a fixed linear power. The development of fission products soluble in the UO 2 phase or, more important, the deliberate introduction of additive oxides in advanced CANDU fuel bundle designs further affects and generally lowers the incipient melting temperature. For these reasons, the modeling of the molten (hyperstoichiometric) UO 2 phase containing several solute oxides (ZrO 2 , Ln 2 O 3 and AnO 2 ) is advancing in the expectation of developing a moving boundary heat and mass transfer model aimed at better defining the limits of safe operating practice as burnup advances. The paper describes how the molten phase stability model is constructed. The redistribution of components across the solid-liquid interface that attends the onset of melting of a non-stoichiometric UO 2 containing several solutes will be discussed. The issues of how to introduce boundary conditions into heat transfer calculations consistent with the requirements of the Phase Rule will be addressed. The Stefan problem of a moving boundary associated with the solid/liquid interface sets this treatment apart from conventional heat and mass transfer problems. (author)

  1. A characterization of Greenland Ice Sheet surface melt and runoff in contemporary reanalyses and a regional climate model

    Directory of Open Access Journals (Sweden)

    Richard eCullather

    2016-02-01

    Full Text Available For the Greenland Ice Sheet (GrIS, large-scale melt area has increased in recent years and is detectable via remote sensing, but its relation to runoff is not known. Historical, modeled melt area and runoff from Modern-Era Retrospective Analysis for Research and Applications (MERRA-Replay, the Interim Re-Analysis of the European Centre for Medium Range Weather Forecasts (ERA-I, the Climate Forecast System Reanalysis (CFSR, the Modèle Atmosphérique Régional (MAR, and the Arctic System Reanalysis (ASR are examined. These sources compare favorably with satellite-derived estimates of surface melt area for the period 2000-2012. Spatially, the models markedly disagree on the number of melt days in the interior of the southern part of the ice sheet, and on the extent of persistent melt areas in the northeastern GrIS. Temporally, the models agree on the mean seasonality of daily surface melt and on the timing of large-scale melt events in 2012. In contrast, the models disagree on the amount, seasonality, spatial distribution, and temporal variability of runoff. As compared to global reanalyses, time series from MAR indicate a lower correlation between runoff and melt area (r2 = 0.805. Runoff in MAR is much larger in the second half of the melt season for all drainage basins, while the ASR indicates larger runoff in the first half of the year. This difference in seasonality for the MAR and to an extent for the ASR provide a hysteresis in the relation between runoff and melt area, which is not found in the other models. The comparison points to a need for reliable observations of surface runoff.

  2. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  3. Seismic risk assessment of a BWR

    International Nuclear Information System (INIS)

    Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

    1987-01-01

    The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

  4. Mathematical modelling of bubble removal from a glass melting channel with defined melt flow and the relation between the optimal flow conditions of bubble removal and sand dissolution

    Czech Academy of Sciences Publication Activity Database

    Cincibusová, Petra; Němec, L.

    2015-01-01

    Roč. 56, č. 2 (2015), s. 52-62 ISSN 1753-3546 R&D Projects: GA TA ČR TA01010844 Institutional support: RVO:67985891 Keywords : glass melt * mathematical modelling * controlled flow * space utilization * temperature gradients Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass Impact factor: 0.362, year: 2015 http://www.ingentaconnect.com/content/sgt/gta/2015/00000056/00000002/art00003

  5. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

    1989-01-01

    BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

  6. Modification of Pawlow's thermodynamical model for the melting of small single-component particles

    Science.gov (United States)

    Barybin, Anatoly; Shapovalov, Victor

    2011-02-01

    A new approach to the melting of small particles is proposed to modify the known Pawlow's model by taking into account the transfer of material from solid spherical particles to liquid ones through a gas phase. Thermodynamical analysis gives rise to a differential equation for the melting point Tm involving such size-dependent and temperature-dependent parameters of a material as the surface tensions σs(l ), molar heat of fusion ΔHm and molar volumes vs(l ). Solution of this equation has shown that all the limiting cases for size-independent situations coincide with results known in the literature and our analysis of size-dependent situations gives results close to the experimental data previously obtained by other authors for some metallic particles.

  7. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    International Nuclear Information System (INIS)

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  8. SCDAP/RELAP5 Modeling of Movement of Melted Material Through Porous Debris in Lower Head

    International Nuclear Information System (INIS)

    Siefken, L. J.

    1998-01-01

    Designs are described for implementing models for calculating the movement of melted material through the interstices in a matrix of porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head during a severe accident in a Light Water Reactor. Currently, the COUPLE model has no capability to model the movement of material that melts within a matrix of porous material. The COUPLE model also does not have the capability to model the movement of liquefied core plate material that slumps onto a porous debris bed in the lower head. In order to advance beyond the assumption the liquefied material always remains stationary, designs are developed for calculations of the movement of liquefied material through the interstices in a matrix of porous material. Correlations are identified for calculating the permeability of the porous debris and for calculating the rate of flow of liquefied material through the interstices in the debris bed. Correlations are also identified for calculating the relocation of solid debris that has a large amount of cavities due to the flowing away of melted material. Equations are defined for calculating the effect on the temperature distribution in the debris bed of heat transported by moving material and for changes in effective thermal conductivity and heat capacity due to the movement of material. The implementation of these models is expected to improve the calculation of the material distribution and temperature distribution of debris in the lower head for cases in which the debris is porous and liquefied material is present within the porous debris

  9. An effective convectivity model for simulation of in-vessel core melt progression in a boiling water reactor

    International Nuclear Information System (INIS)

    Tran, C.T.; Dinh, T.N.

    2007-01-01

    The present paper is concerned with development and application of a so-called Effective Convection Model (ECM), which aims to provide a detailed, mechanistic description of heat transfer processes in a BWR lower plenum. The ECM is a Computational Fluid Dynamics (CFD)-like tool which employs a simpler and more effective approach to compute heat transfer by solving only energy conservation equation instead of solving the full set of Navier-Stokes and energy equations by a CFD code. We implement the ECM in a CFD code (Fluent), with detailed description of the ECM development, implementation and validation. A dual approach is used to validate the ECM, namely validation against experimental data and against heat transfer results obtained by CFD predictions in the same geometries and conditions. Insights gained from CFD simulations are also used to improve ECM. The ECM capability as an effective tool to simulate heat transfer of an internally heated volume in 3-dimensional complex geometry is demonstrated through examples of heat transfer analysis in a BWR lower plenum being cooled by coolant flow in Control Rod Guide Tubes. Simulation results and key findings of this case are reported and discussed. (authors)

  10. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  11. BWR control blade replacement strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

    2000-02-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  12. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  13. Development of next BWR plant

    International Nuclear Information System (INIS)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

    1995-01-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

  14. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    1995-04-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

  15. Modelling snow accumulation and snow melt in a continuous hydrological model for real-time flood forecasting

    International Nuclear Information System (INIS)

    Stanzel, Ph; Haberl, U; Nachtnebel, H P

    2008-01-01

    Hydrological models for flood forecasting in Alpine basins need accurate representation of snow accumulation and snow melt processes. A continuous, semi-distributed rainfall-runoff model with snow modelling procedures using only precipitation and temperature as input is presented. Simulation results from an application in an Alpine Danube tributary watershed are shown and evaluated with snow depth measurements and MODIS remote sensing snow cover information. Seasonal variations of runoff due to snow melt were simulated accurately. Evaluation of simulated snow depth and snow covered area showed strengths and limitations of the model and allowed an assessment of input data quality. MODIS snow cover images were found to be valuable sources of information for hydrological modelling in alpine areas, where ground observations are scarce.

  16. Modelling snow accumulation and snow melt in a continuous hydrological model for real-time flood forecasting

    Energy Technology Data Exchange (ETDEWEB)

    Stanzel, Ph; Haberl, U; Nachtnebel, H P [Institute of Water Management, Hydrology and Hydraulic Engineering, University of Natural Resources and Applied Life Sciences, Muthgasse 18, 1190 Vienna (Austria)], E-mail: philipp.stanzel@boku.ac.at

    2008-11-01

    Hydrological models for flood forecasting in Alpine basins need accurate representation of snow accumulation and snow melt processes. A continuous, semi-distributed rainfall-runoff model with snow modelling procedures using only precipitation and temperature as input is presented. Simulation results from an application in an Alpine Danube tributary watershed are shown and evaluated with snow depth measurements and MODIS remote sensing snow cover information. Seasonal variations of runoff due to snow melt were simulated accurately. Evaluation of simulated snow depth and snow covered area showed strengths and limitations of the model and allowed an assessment of input data quality. MODIS snow cover images were found to be valuable sources of information for hydrological modelling in alpine areas, where ground observations are scarce.

  17. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

    Susilo, Jati

    2002-01-01

    One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6 t h of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

  18. Analytical model based on cohesive energy to indicate the edge and corner effects on melting temperature of metallic nanoparticles

    International Nuclear Information System (INIS)

    Shidpour, Reza; Hamid, Delavari H.; Vossoughi, M.

    2010-01-01

    Graphical abstract: The effect of edge and corner atoms of nanoparticle (solid line) cause melting temperature drops more compared to considering them as same as only surface atoms (dash line). This reduction is significant especially when the size of nanoparticle is below 10 nm. - Abstract: An analytical model based on cohesive energy has been conducted to study the effects of edge, corner, and inward surface relaxation as varying parameters on melting temperature of nanoparticles. It is shown that taking into account the edge and corner (EC) atoms of nanoparticle, causes to drop melting temperature more, when compared to consider them the same as only surface atoms. This reduction is significant especially when the size of nanoparticle is below 10 nm. The results are supported by available experimental results of tin, lead and gold melting temperature (T m ). Finally, it is shown that inward relaxation increases melting temperature slightly.

  19. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

    Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

    1991-01-01

    This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

  20. Modelling of heat transfer between molten core and concrete with account of phase changes in the melt

    International Nuclear Information System (INIS)

    Petukhov, S.M.; Zemlianoukhin, V.V.

    1992-01-01

    The analysis of the process of heat transfer between molten corium and concrete in the case of severe accident in a PWR is performed. It is shown that Bradley's model may be improved for the case of an oxidic melt. A new model is developed and incorporated in the WECHSL-Mod2 Code. Post-test calculations of melt-concrete interaction experiments are carried out. The comparison and analysis of the experimental results and calculations are presented. (9 figures) (Author)

  1. Assessment of the Prony's method for BWR stability analysis

    International Nuclear Information System (INIS)

    Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

    2011-01-01

    Highlights: → This paper describes a method to determine the degree of stability of a BWR. → Performance comparison between Prony's and common AR techniques is presented. → Benchmark data and actual BWR transient data are used for comparison. → DR and f results are presented and discussed. → The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

  2. Property-Composition-Temperature Modeling of Waste Glass Melt Data Subject to a Randomization Restriction

    International Nuclear Information System (INIS)

    Piepel, Gregory F.; Heredia-Langner, Alejandro; Cooley, Scott K.

    2008-01-01

    Properties such as viscosity and electrical conductivity of glass melts are functions of melt temperature as well as glass composition. When measuring such a property for several glasses, the property is typically measured at several temperatures for one glass, then at several temperatures for the next glass, and so on. This data-collection process involves a restriction on randomization, which is referred to as split-plot experiment. The split-plot data structure must be accounted for in developing property-composition-temperature models and the corresponding uncertainty equations for model predictions. Instead of ordinary least squares (OLS) regression methods, generalized least squares (GLS) regression methods using restricted maximum likelihood (REML) estimation must be used. This article describes the methodology for developing property-composition-temperature models and corresponding prediction uncertainty equations using the GLS/REML regression approach. Viscosity data collected on 197 simulated nuclear waste glasses are used to illustrate the GLS/REML methods for developing a viscosity-composition-temperature model and corresponding equations for model prediction uncertainties. The correct results using GLS/REML regression are compared to the incorrect results obtained using OLS regression

  3. Validating predictions made by a thermo-mechanical model of melt segregation in sub-volcanic systems

    Science.gov (United States)

    Roele, Katarina; Jackson, Matthew; Morgan, Joanna

    2014-05-01

    A quantitative understanding of the spatial and temporal evolution of melt distribution in the crust is crucial in providing insights into the development of sub-volcanic crustal stratigraphy and composition. This work aims to relate numerical models that describe the base of volcanic systems with geophysical observations. Recent modelling has shown that the repetitive emplacement of mantle-derived basaltic sills, at the base of the lower crust, acts as a heat source for anatectic melt generation, buoyancy-driven melt segregation and mobilisation. These processes form the lowermost architecture of complex sub-volcanic networks as upward migrating melt produces high melt fraction layers. These 'porosity waves' are separated by zones with high compaction rates and have distinctive polybaric chemical signatures that suggest mixed crust and mantle origins. A thermo-mechanical model produced by Solano et al in 2012 has been used to predict the temperatures and melt fractions of successive high porosity layers within the crust. This model was used as it accounts for the dynamic evolution of melt during segregation and migration through the crust; a significant process that has been neglected in previous models. The results were used to input starting compositions for each of the layers into the rhyolite-MELTS thermodynamic simulation. MELTS then determined the approximate bulk composition of the layers once they had cooled and solidified. The mean seismic wave velocities of the polymineralic layers were then calculated using the relevant Voight-Reuss-Hill mixture rules, whilst accounting for the pressure and temperature dependence of seismic wave velocity. The predicted results were then compared with real examples of reflectivity for areas including the UK, where lower crustal layering is observed. A comparison between the impedance contrasts at compositional boundaries is presented as it confirms the extent to which modelling is able to make predictions that are

  4. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  5. Bond particle model for semiconductor melts and its application to liquid structure germanium

    International Nuclear Information System (INIS)

    Ferrante, A.; Tosi, M.P.

    1988-08-01

    A simple type of liquid state model is proposed to describe on a primitive level the melt of an elemental group IV semiconductor as a mixture of atoms and bond particles. The latter, on increase of a coupling strength parameter becomes increasingly localized between pairs of atoms up to local tetrahedral coordination of atoms by bond particles. Angular interatomic correlations are built into the model as bond particle localization grows, even though the bare interactions between the components of the liquid are formally described solely in terms of central pair potentials. The model is solved for liquid structure by standard integral equation techniques of liquid state theory and by Monte Carlo simulation, for values of the parameters which are appropriate to liquid germanium down to strongly supercooled states. The calculated liquid structure is compared with the results of diffraction experiments on liquid germanium near freezing and discussed in relation to diffraction data on amorphous germanium. The model suggests simple melting criteria for elemental and polar semiconductors, which are empirically verified. (author). 25 refs, 9 figs, 3 tabs

  6. Study on thermal performance and margins of BWR fuel elements

    International Nuclear Information System (INIS)

    Stosic, Zoran

    1999-01-01

    This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

  7. Modelling of RPV lower head under core melt severe accident condition using OpenFOAM

    International Nuclear Information System (INIS)

    Madokoro, Hiroshi; Kretzschmar, Frank; Miassoedov, Alexei

    2017-01-01

    Although six years have been passed since the tragic severe accident at Fukushima Daiichi, still large uncertainties exist in modeling of core degradation and reactor pressure vessel (RPV) failure. It is extremely important to obtain a better understanding of complex phenomena in the lower head in order to improve accident management measures. The possible failure mode of reactor pressure vessel and its failure time are especially a matter of importance. Thermal behavior of the molten pool can be simulated by the Phase-change Effective Convectivity Model (PECM), which is a distributed-parameter model developed in the Royal Institute of Technology (KTH), Sweden. The model calculates convective currents not using a pure CFD approach but based on so called “characteristic velocities” that are determined by empirical correlations depending on the geometry and physical properties of the molten pool. At the Karlsruhe Institute of Technology (KIT), the PECM has been implemented in the open-source CFD software OpenFOAM in order to receive detailed predictions of a core melt behavior in the RPV lower head under severe accident conditions. An advantage of using OpenFOAM is that it is very flexible to add and modify models and physical properties. In the current work, the solver is extended to couple PECM with a structure analysis model of the vessel wall. The model considers thermal expansion, plasticity, creep and damage. The model and physical properties are based on those implemented in ANSYS. Although the previous implementation had restriction that the amount of and geometry of the melt cannot be changed, our coupled model allows flexibility of the melt amount and geometry. The extended solver was used to simulate the LIVE-L1 and -L7V experiments and has demonstrated good prediction of the temperature distribution in the molten pool and heat flux distribution through the vessel wall. Regarding the vessel failure the model was applied to one of the FOREVER tests

  8. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  9. Evaluation of internal flooding in a BWR

    International Nuclear Information System (INIS)

    Shiu, K.; Papazoglou, I.A.; Sun, Y.H.; Anavim, E.; Ilberg, D.

    1985-01-01

    Flooding inside a nuclear power station is capable of concurrently disabling redundant safety systems. This paper presents the results of a recent review study performed on internally-generated floods inside a boiling water reactor (BWR) reactor building. The study evaluated the flood initiator frequency due to either maintenance or ruptures using Markovian models. A time phased event tree approach was adopted to quantify the core damage frequency based on the flood initiator frequency. It is found in the study that the contribution to the total core damage due to internal flooding events is not insignificant and is comparable to other transient contributors. The findings also indicate that the operator plays an important role in the prevention as well as the mitigation of a flooding event

  10. A conceptual model for the asthenosphere: redox melting in the C-O-H-bearing mantle vs. geophysical observations

    Science.gov (United States)

    Gaillard, Fabrice; Tarits, Pascal; Massuyeau, Malcolm; David, Sifre; Leila, Hashim; Emmanuel, Gardes

    2013-04-01

    pump results in an increasingly reduced mantle with depth. Recent surveys have calibrated the carbon-carbonate redox transition at mantle pressure and have located its depth around 180-250 km (depth of redox melting); Deeper, only diamond is stable; Shallower, carbonates, mostly in its molten state, are expected. This petrological model is confronted to the most recent geophysical observations. Such observations indicate that melting must occur at depth down to 400 km, which conflict with the concept of redox melting. What is the composition of the melt? Hydrous silicate melt or hydrous carbonated melt? What does it mean in terms of deep upper mantle redox state?

  11. Hydraulic modeling and simulation of a System Division of Essential Service Water in a BWR plant with Flow master; Modelo hidraulico y simulacion de una division del Sistema de Agua de Servicio Esencial de una central BWR con Flowmaster

    Energy Technology Data Exchange (ETDEWEB)

    Vegazo Juzgado, L.; Rodriguez Garcia, G. M.; Mota Coloma, M.

    2012-07-01

    At the conclusion of the project can say that Flow master is a simulation tool that allows you to create your model from a library of components and obtain useful results from the point of view of the operation, engineering and maintenance. Compared to previous software from the point of view of use, can comment that Flow master is a tool which has an intuitive and user-friendly interaction between the user and the program thus facilitating the modeling of the system and definition of the components of same.

  12. The 1988-2003 Greenland ice sheet melt extent using passive microwave satellite data and a regional climate model

    Energy Technology Data Exchange (ETDEWEB)

    Fettweis, Xavier; Ypersele, Jean-Pascal van [Universite Catholique de Louvain, Institut d' Astronomie et de Geophysique de G. Lemaitre, Louvain-La-Neuve (Belgium); Gallee, Hubert [CNRS, Laboratoire de Glaciologie et Geophysique de l' Environnement, Grenoble (France); Lefebre, Filip [Vito-IMS (Flemish Institute for Technological Research-Integral Environmental Studies), Mol (Belgium)

    2006-10-15

    Measurements from ETH-Camp and JAR1 AWS (West Greenland) as well as coupled atmosphere-snow regional climate simulations have highlighted flaws in the cross-polarized gradient ratio (XPGR) technique used to identify melt from passive microwave satellite data. It was found that dense clouds (causing notably rainfall) on the ice sheet severely perturb the XPGR melt signal. Therefore, the original XPGR melt detection algorithm has been adapted to better incorporate atmospheric variability over the ice sheet and an updated melt trend for the 1988-2003 period has been calculated. Compared to the original algorithm, the melt zone area increase is eight times higher (from 0.2 to 1.7% year{sup -1}). The increase is higher with the improved XPGR technique because rainfall also increased during this period. It is correlated to higher atmospheric temperatures. Finally, the model shows that the total ice sheet runoff is directly proportional to the melt extent surface detected by satellites. These results are important for the understanding of the effect of Greenland melting on the stability of the thermohaline circulation. (orig.)

  13. Finite element modeling of melting and fluid flow in the laser-heated diamond-anvil cell

    Science.gov (United States)

    Gomez-Perez, N.; Rodriguez, J. F.; McWilliams, R. S.

    2017-04-01

    The laser-heated diamond anvil cell is widely used in the laboratory study of materials behavior at high-pressure and high-temperature, including melting curves and liquid properties at extreme conditions. Laser heating in the diamond cell has long been associated with fluid-like motion in samples, which is routinely used to determine melting points and is often described as convective in appearance. However, the flow behavior of this system is poorly understood. A quantitative treatment of melting and flow in the laser-heated diamond anvil cell is developed here to physically relate experimental motion to properties of interest, including melting points and viscosity. Numerical finite-element models are used to characterize the temperature distribution, melting, buoyancy, and resulting natural convection in samples. We find that continuous fluid motion in experiments can be explained most readily by natural convection. Fluid velocities, peaking near values of microns per second for plausible viscosities, are sufficiently fast to be detected experimentally, lending support to the use of convective motion as a criterion for melting. Convection depends on the physical properties of the melt and the sample geometry and is too sluggish to detect for viscosities significantly above that of water at ambient conditions, implying an upper bound on the melt viscosity of about 1 mPa s when convective motion is detected. A simple analytical relationship between melt viscosity and velocity suggests that direct viscosity measurements can be made from flow speeds, given the basic thermodynamic and geometric parameters of samples are known.

  14. A genetic model of progressively partial melting for uranium-bearing granites in south China

    International Nuclear Information System (INIS)

    Zhai Jianping.

    1989-01-01

    A genetic model of progressively partial and enrichment mechanism of uranium during partial melting of the sources of material studied and the significance of the genetic model in search of uranium deposits is elaborated. This model accounts better for some geological and geochemical features of uranium-bearing granties and suspects the traditional idea that igneous uranium-bearing granites were formed by fusion of U-rich strata surrounding these granites. Finally this paper points out that the infuence of U-rich strata of wall rocks of granites over uranium-bearing granites depends on variation of water solubility in the magma and assimilation of magma to wall rocks during its ascending and crystallization

  15. Modeling of Melting and Resolidification in Domain of Metal Film Subjected to a Laser Pulse

    Directory of Open Access Journals (Sweden)

    Majchrzak E.

    2016-03-01

    Full Text Available Thermal processes in domain of thin metal film subjected to a strong laser pulse are discussed. The heating of domain considered causes the melting and next (after the end of beam impact the resolidification of metal superficial layer. The laser action (a time dependent bell-type function is taken into account by the introduction of internal heat source in the energy equation describing the heat transfer in domain of metal film. Taking into account the extremely short duration, extreme temperature gradients and very small geometrical dimensions of the domain considered, the mathematical model of the process is based on the dual phase lag equation supplemented by the suitable boundary-initial conditions. To model the phase transitions the artificial mushy zone is introduced. At the stage of numerical modeling the Control Volume Method is used. The examples of computations are also presented.

  16. Spatio-temporal Variability of Albedo and its Impact on Glacier Melt Modelling

    Science.gov (United States)

    Kinnard, C.; Mendoza, C.; Abermann, J.; Petlicki, M.; MacDonell, S.; Urrutia, R.

    2017-12-01

    Albedo is an important variable for the surface energy balance of glaciers, yet its representation within distributed glacier mass-balance models is often greatly simplified. Here we study the spatio-temporal evolution of albedo on Glacier Universidad, central Chile (34°S, 70°W), using time-lapse terrestrial photography, and investigate its effect on the shortwave radiation balance and modelled melt rates. A 12 megapixel digital single-lens reflex camera was setup overlooking the glacier and programmed to take three daily images of the glacier during a two-year period (2012-2014). One image was chosen for each day with no cloud shading on the glacier. The RAW images were projected onto a 10m resolution digital elevation model (DEM), using the IMGRAFT software (Messerli and Grinsted, 2015). A six-parameter camera model was calibrated using a single image and a set of 17 ground control points (GCPs), yielding a georeferencing accuracy of accounting for possible camera movement over time. The reflectance values from the projected image were corrected for topographic and atmospheric influences using a parametric solar irradiation model, following a modified algorithm based on Corripio (2004), and then converted to albedo using reference albedo measurements from an on-glacier automatic weather station (AWS). The image-based albedo was found to compare well with independent albedo observations from a second AWS in the glacier accumulation area. Analysis of the albedo maps showed that the albedo is more spatially-variable than the incoming solar radiation, making albedo a more important factor of energy balance spatial variability. The incorporation of albedo maps within an enhanced temperature index melt model revealed that the spatio-temporal variability of albedo is an important factor for the calculation of glacier-wide meltwater fluxes.

  17. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

    1990-03-01

    Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

  18. Seismic PRA of a BWR plant

    International Nuclear Information System (INIS)

    Nishio, Masahide; Fujimoto, Haruo

    2014-01-01

    Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

  19. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator; Modelado de la dinamica de la vasija y circuitos de recirculacion de una nucleoelectrica tipo BWR como parte del simulador universitario SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R A [DEPFI, Campus Morelos, en IMTA, Jiutepec, Morelos (Mexico)

    2003-07-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  20. Modeling the heating and melting of sea ice through light absorption by microalgae

    Science.gov (United States)

    Zeebe, Richard E.; Eicken, Hajo; Robinson, Dale H.; Wolf-Gladrow, Dieter; Dieckmann, Gerhard S.

    1996-01-01

    In sea ice of polar regions, high concentrations of microalgae are observed during the spring. Algal standing stocks may attain peak values of over 300 mg chl a m-2 in the congelation ice habitat. As of yet, the effect of additional heating of sea ice through conversion of solar radiation into heat by algae has not been investigated in detail. Local effects, such as a decrease in albedo, increasing melt rates, and a decrease of the physical strength of ice sheets may occur. To investigate the effects of microalgae on the thermal regime of sea ice, a time-dependent, one-dimensional thermodynamic model of sea ice was coupled to a bio-optical model. A spectral one-stream model was employed to determine spectral attenuation by snow, sea ice, and microalgae. Beer's law was assumed to hold for every wavelength. Energy absorption was obtained by calculating the divergence of irradiance in every layer of the model (Δz = 1 cm). Changes in sea ice temperature profiles were calculated by solving the heat conduction equation with a finite difference scheme. Model results indicate that when algal biomass is concentrated at the bottom of congelation ice, melting of ice resulting from the additional conversion of solar radiation into heat may effectively destroy the algal habitat, thereby releasing algal biomass into the water column. An algal layer located in the top of the ice sheet induced a significant increase in sea ice temperature (ΔT > 0.3 K) for snow depths less than 5 cm and algal standing stocks higher than 150 mg chl a m-2. Furthermore, under these conditions, brine volume increased by 21% from 181 to 219 parts per thousand, which decreased the physical strength of the ice.

  1. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

    1992-01-01

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  2. Strength analysis and modeling of cellular lattice structures manufactured using selective laser melting for tooling applications

    DEFF Research Database (Denmark)

    Mahshid, Rasoul; Hansen, Hans Nørgaard; Loft Højbjerre, Klaus

    2016-01-01

    Additive manufacturing is rapidly developing and gaining popularity for direct metal fabrication systems like selective laser melting (SLM). The technology has shown significant improvement for high-quality fabrication of lightweight design-efficient structures such as conformal cooling channels...... in injection molding tools and lattice structures. This research examines the effect of cellular lattice structures on the strength of workpieces additively manufactured from ultra high-strength steel powder. Two commercial SLM machines are used to fabricate cellular samples based on four architectures— solid......, hollow, lattice structure and rotated lattice structure. Compression test is applied to the specimens while they are deformed. The analytical approach includes finite element (FE), geometrical and mathematical models for prediction of collapse strength. The results from the the models are verified...

  3. Modelling of pulsed electron beam induced graphite ablation: Sublimation versus melting

    Science.gov (United States)

    Ali, Muddassir; Henda, Redhouane

    2017-12-01

    Pulsed electron beam ablation (PEBA) has recently emerged as a very promising technique for the deposition of thin films with superior properties. Interaction of the pulsed electron beam with the target material is a complex process, which consists of heating, phase transition, and erosion of a small portion from the target surface. Ablation can be significantly affected by the nature of thermal phenomena taking place at the target surface, with subsequent bearing on the properties, stoichiometry and structure of deposited thin films. A two stage, one-dimensional heat conduction model is presented to describe two different thermal phenomena accounting for interaction of a graphite target with a polyenergetic electron beam. In the first instance, the thermal phenomena are comprised of heating, melting and vaporization of the target surface, while in the second instance the thermal phenomena are described in terms of heating and sublimation of the graphite surface. In this work, the electron beam delivers intense electron pulses of ∼100 ns with energies up to 16 keV and an electric current of ∼400 A to a graphite target. The temperature distribution, surface recession velocity, ablated mass per unit area, and ablation depth for the graphite target are numerically simulated by the finite element method for each case. Based on calculation findings and available experimental data, ablation appears to occur mainly in the regime of melting and vaporization from the surface.

  4. Degenerate Ising model for atomistic simulation of crystal-melt interfaces

    International Nuclear Information System (INIS)

    Schebarchov, D.; Schulze, T. P.; Hendy, S. C.

    2014-01-01

    One of the simplest microscopic models for a thermally driven first-order phase transition is an Ising-type lattice system with nearest-neighbour interactions, an external field, and a degeneracy parameter. The underlying lattice and the interaction coupling constant control the anisotropic energy of the phase boundary, the field strength represents the bulk latent heat, and the degeneracy quantifies the difference in communal entropy between the two phases. We simulate the (stochastic) evolution of this minimal model by applying rejection-free canonical and microcanonical Monte Carlo algorithms, and we obtain caloric curves and heat capacity plots for square (2D) and face-centred cubic (3D) lattices with periodic boundary conditions. Since the model admits precise adjustment of bulk latent heat and communal entropy, neither of which affect the interface properties, we are able to tune the crystal nucleation barriers at a fixed degree of undercooling and verify a dimension-dependent scaling expected from classical nucleation theory. We also analyse the equilibrium crystal-melt coexistence in the microcanonical ensemble, where we detect negative heat capacities and find that this phenomenon is more pronounced when the interface is the dominant contributor to the total entropy. The negative branch of the heat capacity appears smooth only when the equilibrium interface-area-to-volume ratio is not constant but varies smoothly with the excitation energy. Finally, we simulate microcanonical crystal nucleation and subsequent relaxation to an equilibrium Wulff shape, demonstrating the model's utility in tracking crystal-melt interfaces at the atomistic level

  5. Degenerate Ising model for atomistic simulation of crystal-melt interfaces

    Energy Technology Data Exchange (ETDEWEB)

    Schebarchov, D., E-mail: Dmitri.Schebarchov@gmail.com [University Chemical Laboratories, Lensfield Road, Cambridge CB2 1EW (United Kingdom); Schulze, T. P., E-mail: schulze@math.utk.edu [Department of Mathematics, University of Tennessee, Knoxville, Tennessee 37996-1300 (United States); Hendy, S. C. [The MacDiarmid Institute for Advanced Materials and Nanotechnology, School of Chemical and Physical Sciences, Victoria University of Wellington, Wellington 6140 (New Zealand); Department of Physics, University of Auckland, Auckland 1010 (New Zealand)

    2014-02-21

    One of the simplest microscopic models for a thermally driven first-order phase transition is an Ising-type lattice system with nearest-neighbour interactions, an external field, and a degeneracy parameter. The underlying lattice and the interaction coupling constant control the anisotropic energy of the phase boundary, the field strength represents the bulk latent heat, and the degeneracy quantifies the difference in communal entropy between the two phases. We simulate the (stochastic) evolution of this minimal model by applying rejection-free canonical and microcanonical Monte Carlo algorithms, and we obtain caloric curves and heat capacity plots for square (2D) and face-centred cubic (3D) lattices with periodic boundary conditions. Since the model admits precise adjustment of bulk latent heat and communal entropy, neither of which affect the interface properties, we are able to tune the crystal nucleation barriers at a fixed degree of undercooling and verify a dimension-dependent scaling expected from classical nucleation theory. We also analyse the equilibrium crystal-melt coexistence in the microcanonical ensemble, where we detect negative heat capacities and find that this phenomenon is more pronounced when the interface is the dominant contributor to the total entropy. The negative branch of the heat capacity appears smooth only when the equilibrium interface-area-to-volume ratio is not constant but varies smoothly with the excitation energy. Finally, we simulate microcanonical crystal nucleation and subsequent relaxation to an equilibrium Wulff shape, demonstrating the model's utility in tracking crystal-melt interfaces at the atomistic level.

  6. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Porsmyr, Jan; Bodal, Terje; Beere, William H.

    2004-01-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

  7. SCORPIO-BWR: status and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

    2004-07-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

  8. LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

    International Nuclear Information System (INIS)

    Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

    2005-01-01

    Full text of publication follows: Unstable oscillation of a nuclear power reactor core is one of the main reasons that causes minor core damage. Stability analysis needs to be performed to predict the potential problem as early as possible and to prevent core instability events from happening. Nuclear Regulatory Commission (NRC) requests all BWR licensees to examine each core reload and to impose operating limitations, as appropriate, to ensure compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. Therefore, the core instability is directly related to the fuel design limits. The core and channel DR (decay ratio) calculation are commonly performed to determine system's stability when new fuel designs are introduced in the core. In order to establish the independent analysis technology for BWR licensees and verifications, the Institute of Nuclear Energy Research (INER) has obtained agreement from NRC and implemented the 'Methodology and Procedure for Calculation of Core and Channel Decay Ratios with LAPUR', which was developed by the IBERINCO in 2001. LAPUR5 uses a multi-nodal description of the neutron dynamics, together with a distributed parameter model of the core thermal hydrodynamics to produce a space-dependent representation of the dynamics of a BWR in the frequency domain for small perturbations around a steady state condition. From the output of LAPUR5, the following results are obtained: global core decay ratio, out-of phase core decay ratio, and channel decay ratio. They are key parameters in the determination of BWR core stability

  9. CADASTER QSPR Models for Predictions of Melting and Boiling Points of Perfluorinated Chemicals.

    Science.gov (United States)

    Bhhatarai, Barun; Teetz, Wolfram; Liu, Tao; Öberg, Tomas; Jeliazkova, Nina; Kochev, Nikolay; Pukalov, Ognyan; Tetko, Igor V; Kovarich, Simona; Papa, Ester; Gramatica, Paola

    2011-03-14

    Quantitative structure property relationship (QSPR) studies on per- and polyfluorinated chemicals (PFCs) on melting point (MP) and boiling point (BP) are presented. The training and prediction chemicals used for developing and validating the models were selected from Syracuse PhysProp database and literatures. The available experimental data sets were split in two different ways: a) random selection on response value, and b) structural similarity verified by self-organizing-map (SOM), in order to propose reliable predictive models, developed only on the training sets and externally verified on the prediction sets. Individual linear and non-linear approaches based models developed by different CADASTER partners on 0D-2D Dragon descriptors, E-state descriptors and fragment based descriptors as well as consensus model and their predictions are presented. In addition, the predictive performance of the developed models was verified on a blind external validation set (EV-set) prepared using PERFORCE database on 15 MP and 25 BP data respectively. This database contains only long chain perfluoro-alkylated chemicals, particularly monitored by regulatory agencies like US-EPA and EU-REACH. QSPR models with internal and external validation on two different external prediction/validation sets and study of applicability-domain highlighting the robustness and high accuracy of the models are discussed. Finally, MPs for additional 303 PFCs and BPs for 271 PFCs were predicted for which experimental measurements are unknown. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  10. Specifications of the BWR simulator for HAMMLAB 2000

    International Nuclear Information System (INIS)

    Grini, Rolf-Einar; Miettinen, Jaakko; Nurmilaukas, Pekka; Raussi; Pekka; Saarni, Ray; Stokke; Egil; Soerensen, Aimar; Tiihonen, Olli

    1998-02-01

    The Boiling Water Reactor (BWR) simulator for HAMMLAB 2000 will be a model of the Swedish plant Forsmark-3. This report gives the specifications of the BWR simulator. The bulk of the report is a copy of the relevant addendum to the contract with the developer, and to the contract with the group of utilities and with ABB Atom. After a general overview, each plant system is described one after the other (using the reference plant system coding), and the simulation of each system is specified. Even the systems that shall not be simulated are included; in those cases the specification is: It is not required that ... is simulated. A list of malfunctions is given, as well as a list of validation transients. Finally the operator interface is specified. (author)

  11. Waveform modeling of the seismic response of a mid-ocean ridge axial melt sill

    Science.gov (United States)

    Xu, Min; Stephen, R. A.; Canales, J. Pablo

    2017-12-01

    Seismic reflections from axial magma lens (AML) are commonly observed along many mid-ocean ridges, and are thought to arise from the negative impedance contrast between a solid, high-speed lid and the underlying low-speed, molten or partially molten (mush) sill. The polarity of the AML reflection ( P AML P) at vertical incidence and the amplitude vs offset (AVO) behavior of the AML reflections (e.g., P AML P and S-converted P AML S waves) are often used as a diagnostic tool for the nature of the low-speed sill. Time-domain finite difference calculations for two-dimensional laterally homogeneous models show some scenarios make the interpretation of melt content from partial-offset stacks of P- and S-waves difficult. Laterally heterogeneous model calculations indicate diffractions from the edges of the finite-width AML reducing the amplitude of the AML reflections. Rough seafloor and/or a rough AML surface can also greatly reduce the amplitude of peg-leg multiples because of scattering and destructive interference. Mid-crustal seismic reflection events are observed in the three-dimensional multi-channel seismic dataset acquired over the RIDGE-2000 Integrated Study Site at East Pacific Rise (EPR, cruise MGL0812). Modeling indicates that the mid-crustal seismic reflection reflections are unlikely to arise from peg-leg multiples of the AML reflections, P-to- S converted phases, or scattering due to rough topography, but could probably arise from deeper multiple magma sills. Our results support the identification of Marjanović et al. (Nat Geosci 7(11):825-829, 2014) that a multi-level complex of melt lenses is present beneath the axis of the EPR.

  12. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

    2012-09-01

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  13. Modelling present-day basal melt rates for Antarctic ice shelves using a parametrization of buoyant meltwater plumes

    Science.gov (United States)

    Lazeroms, Werner M. J.; Jenkins, Adrian; Hilmar Gudmundsson, G.; van de Wal, Roderik S. W.

    2018-01-01

    Basal melting below ice shelves is a major factor in mass loss from the Antarctic Ice Sheet, which can contribute significantly to possible future sea-level rise. Therefore, it is important to have an adequate description of the basal melt rates for use in ice-dynamical models. Most current ice models use rather simple parametrizations based on the local balance of heat between ice and ocean. In this work, however, we use a recently derived parametrization of the melt rates based on a buoyant meltwater plume travelling upward beneath an ice shelf. This plume parametrization combines a non-linear ocean temperature sensitivity with an inherent geometry dependence, which is mainly described by the grounding-line depth and the local slope of the ice-shelf base. For the first time, this type of parametrization is evaluated on a two-dimensional grid covering the entire Antarctic continent. In order to apply the essentially one-dimensional parametrization to realistic ice-shelf geometries, we present an algorithm that determines effective values for the grounding-line depth and basal slope in any point beneath an ice shelf. Furthermore, since detailed knowledge of temperatures and circulation patterns in the ice-shelf cavities is sparse or absent, we construct an effective ocean temperature field from observational data with the purpose of matching (area-averaged) melt rates from the model with observed present-day melt rates. Our results qualitatively replicate large-scale observed features in basal melt rates around Antarctica, not only in terms of average values, but also in terms of the spatial pattern, with high melt rates typically occurring near the grounding line. The plume parametrization and the effective temperature field presented here are therefore promising tools for future simulations of the Antarctic Ice Sheet requiring a more realistic oceanic forcing.

  14. Point, surface and volumetric heat sources in the thermal modelling of selective laser melting

    Science.gov (United States)

    Yang, Yabin; Ayas, Can

    2017-10-01

    Selective laser melting (SLM) is a powder based additive manufacturing technique suitable for producing high precision metal parts. However, distortions and residual stresses within products arise during SLM because of the high temperature gradients created by the laser heating. Residual stresses limit the load resistance of the product and may even lead to fracture during the built process. It is therefore of paramount importance to predict the level of part distortion and residual stress as a function of SLM process parameters which requires a reliable thermal modelling of the SLM process. Consequently, a key question arises which is how to describe the laser source appropriately. Reasonable simplification of the laser representation is crucial for the computational efficiency of the thermal model of the SLM process. In this paper, first a semi-analytical thermal modelling approach is described. Subsequently, the laser heating is modelled using point, surface and volumetric sources, in order to compare the influence of different laser source geometries on the thermal history prediction of the thermal model. The present work provides guidelines on appropriate representation of the laser source in the thermal modelling of the SLM process.

  15. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

    2006-01-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde

  16. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  17. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  18. Incorporation of a physically based melt pond scheme into the sea ice component of a climate model

    OpenAIRE

    Flocco, Daniela; Feltham, Danny; Turner, Adrian K.

    2010-01-01

    The extent and thickness of the Arctic sea ice cover has decreased dramatically in the past few decades with minima in sea ice extent in September 2005 and 2007. These minima have not been predicted in the IPCC AR4 report, suggesting that the sea ice component of climate models should more realistically represent the processes controlling the sea ice mass balance. One of the processes poorly represented in sea ice models is the formation and evolution of melt ponds. Melt ponds accumulate on t...

  19. Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

    2006-07-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

  20. Modelling the geometry of a moving laser melt pool and deposition track via energy and mass balances

    Energy Technology Data Exchange (ETDEWEB)

    Pinkerton, Andrew J; Li Lin [Laser Processing Research Centre, Department of Mechanical, Aerospace and Manufacturing Engineering, University of Manchester Institute of Science and Technology, PO Box 88, Sackville Street, Manchester M60 1QD (United Kingdom)

    2004-07-21

    The additive manufacturing technique of laser direct metal deposition allows multiple tracks of full density metallic material to be built to form complex parts for rapid tooling and manufacture. Practical results and theoretical models have shown that the geometries of the tracks are governed by multiple factors. Original work with single layer cladding identified three basic clad profiles but, so far, models of multiple layer, powder-feed deposition have been based on only two of them. At higher powder mass flow rates, experimental results have shown that a layer's width can become greater than the melt pool width at the substrate surface, but previous analytical models have not been able to accommodate this. In this paper, a model based on this third profile is established and experimentally verified. The model concentrates on mathematical analysis of the melt pool and establishes mass and energy balances based on one-dimensional heat conduction to the substrate. Deposition track limits are considered as arcs of circles rather than of ellipses, as used in most established models, reflecting the dominance of surface tension forces in the melt pool, and expressions for elongation of the melt pool with increasing traverse speed are incorporated. Trends in layer width and height with major process parameters are captured and predicted layer dimensions correspond well to the experimental values.

  1. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  2. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  3. Using an Ablation Gradient Model to Characterize Annual Glacial Melt Contribution to Major Rivers in High Asia

    Science.gov (United States)

    Brodzik, M. J.; Armstrong, R. L.; Khalsa, S. J. S.; Painter, T. H.; Racoviteanu, A.; Rittger, K.

    2014-12-01

    Ice melt from mountain glaciers can represent a significant contribution to freshwater hydrological budgets, along with seasonal snow melt, rainfall and groundwater. In the rivers of High Asia, understanding the proportion of glacier ice melt is critical for water resource management of irrigation and planning for hydropower generation and human consumption. Current climate conditions are producing heterogeneous glacier responses across the Hindu Kush-Karakoram-Himalayan ranges. However, it is not yet clear how contrasting glacier patterns affect regional water resources. For example, in the Upper Indus basin, estimates of glacial contribution to runoff are often not distinguished from seasonal snow contribution, and vary widely, from as little as 15% to as much as 55%. While many studies are based on reasonable concepts, most are based on assumptions uninformed by actual snow or ice cover measurements. While straightforward temperature index models have been used to estimate glacier runoff in some Himalayan basins, application of these models in larger Himalayan basins is limited by difficulties in estimating key model parameters, particularly air temperature. Estimating glacial area from the MODIS Permanent Snow and Ice Extent (MODICE) product for the years 2000-2013, with recently released Shuttle Radar Topography Mission (SRTMGL3) elevation data, we use a simple ablation gradient approach to calculate an upper limit on the contribution of clean glacier ice melt to streamflow data. We present model results for the five major rivers with glaciated headwaters in High Asia: the Bramaputra, Ganges, Indus, Amu Darya and Syr Darya. Using GRDC historical discharge records, we characterize the annual contribution from glacier ice melt. We use MODICE interannual trends in each basin to estimate glacier ice melt uncertainties. Our results are being used in the USAID project, Contribution to High Asia Runoff from Ice and Snow (CHARIS), to inform regional-scale planning for

  4. Kinetic model for quartz and spinel dissolution during melting of high-level-waste glass batch

    International Nuclear Information System (INIS)

    Pokorny, Richard; Rice, Jarrett A.; Crum, Jarrod V.; Schweiger, Michael J.; Hrma, Pavel

    2013-01-01

    The dissolution of quartz particles and the growth and dissolution of crystalline phases during the conversion of batch to glass potentially affects both the glass melting process and product quality. Crystals of spinel exiting the cold cap to molten glass below can be troublesome during the vitrification of iron-containing high-level wastes. To estimate the distribution of quartz and spinel fractions within the cold cap, we used kinetic models that relate fractions of these phases to temperature and heating rate. Fitting the model equations to data showed that the heating rate, apart from affecting quartz and spinel behavior directly, also affects them indirectly via concurrent processes, such as the formation and motion of bubbles. Because of these indirect effects, it was necessary to allow one kinetic parameter (the pre-exponential factor) to vary with the heating rate. The resulting kinetic equations are sufficiently simple for the detailed modeling of batch-to-glass conversion as it occurs in glass melters. The estimated fractions and sizes of quartz and spinel particles as they leave the cold cap, determined in this study, will provide the source terms needed for modeling the behavior of these solid particles within the flow of molten glass in the melter

  5. A flow time model for melt-cast insensitive explosive process

    Energy Technology Data Exchange (ETDEWEB)

    Guillemin, Jean-Philippe; Brunet, Luc [Nexter Munitions, 7 Route de Guerry, 18023 Bourges Cedex (France); Bonnefoy, Olivier; Thomas, Gerard [Ecole Nationale Superieure des Mines de Saint-Etienne, Centre SPIN/LPMG, UMR CNRS 5148, 158 Cours Fauriel, 42023 Saint-Etienne Cedex 2 (France)

    2007-06-15

    Diphasic flows of concentrated suspensions of melt-cast insensitive explosives exhibit specific rheological properties. In order to limit the handling of pyrotechnical products presenting a risk with respect to the mechanical and thermal shocks, a lot of work has been undertaken for many years in the civil engineering sector. The objective of this study is to propose a predictive model of the flow time of a concentrated suspension through a nozzle located at the bottom of a tank. Similar to our industrial process, the suspension is made out of insensitive energetic materials and flows under gravity. Experimental results are compared to three models (Quemada, Krieger-Dougherty, and Mooney) predicting the viscosity {mu} of a suspension as a function of the solid volume fraction {phi}, the maximum packing density {phi}{sub m} and the viscosity {mu}{sub 0} of the interstitial liquid. De Larrard's model is used to calculate {phi}{sub m}. The value of viscosity measured for the pure liquid is close to the one predicted by the Bernoulli theorem, where liquids are considered as incompressible and inviscid. Finally, it was found that the Quemada's model gives a fair agreement between predictions and experiments. (Abstract Copyright [2007], Wiley Periodicals, Inc.)

  6. Implementation of a thermomechanical model for the simulation of selective laser melting

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, N. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, R. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, J. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-07-01

    Selective laser melting (SLM) is an additive manufacturing process in which multiple, successive layers of metal powders are heated via laser in order to build a part. Modeling of SLM requires consideration of both heat transfer and solid mechanics. The present work describes continuum modeling of SLM as envisioned for eventual support of part-scale modeling of this fabrication process to determine end-state information such as residual stresses and distortion. The determination of the evolving temperatures is dependent on the material, the state of the material (powder or solid), the specified heating, and the configuration. Similarly, the current configuration is dependent on the temperatures, the powder-solid state, and the constitutive models. A multi-physics numerical formulation is required to solve such problems. This article describes the problem formulation, numerical method, and constitutive parameters necessary to solve such a problem. Additionally, various verification and example problems are simulated in the parallel, multi-physics finite element code Diablo, and the results presented herein.

  7. Modelling the evaporation of boron species. Part 1: Alkali-free borosilicate glass melts

    NARCIS (Netherlands)

    Limpt, J.A.C. van; Beerkens, R.G.C.; Cook, S.; O'Connor, R.; Simon, J.

    2011-01-01

    A laboratory test facility has been used to measure the boron evaporation rates from borosilicate glass melts. The impact of furnace atmosphere composition and glass melt composition on the temperature dependent boron evaporation rates has been investigated experimentally. In Part 1 of this paper

  8. Meteorological conditions associated to high sublimation amounts in semiarid high-elevation Andes decrease the performance of empirical melt models

    Science.gov (United States)

    Ayala, Alvaro; Pellicciotti, Francesca; MacDonell, Shelley; McPhee, James; Burlando, Paolo

    2015-04-01

    Empirical melt (EM) models are often preferred to surface energy balance (SEB) models to calculate melt amounts of snow and ice in hydrological modelling of high-elevation catchments. The most common reasons to support this decision are that, in comparison to SEB models, EM models require lower levels of meteorological data, complexity and computational costs. However, EM models assume that melt can be characterized by means of a few index variables only, and their results strongly depend on the transferability in space and time of the calibrated empirical parameters. In addition, they are intrinsically limited in accounting for specific process components, the complexity of which cannot be easily reconciled with the empirical nature of the model. As an example of an EM model, in this study we use the Enhanced Temperature Index (ETI) model, which calculates melt amounts using air temperature and the shortwave radiation balance as index variables. We evaluate the performance of the ETI model on dry high-elevation sites where sublimation amounts - that are not explicitly accounted for the EM model - represent a relevant percentage of total ablation (1.1 to 8.7%). We analyse a data set of four Automatic Weather Stations (AWS), which were collected during the ablation season 2013-14, at elevations between 3466 and 4775 m asl, on the glaciers El Tapado, San Francisco, Bello and El Yeso, which are located in the semiarid Andes of central Chile. We complement our analysis using data from past studies in Juncal Norte Glacier (Chile) and Haut Glacier d'Arolla (Switzerland), during the ablation seasons 2008-09 and 2006, respectively. We use the results of a SEB model, applied to each study site, along the entire season, to calibrate the ETI model. The ETI model was not designed to calculate sublimation amounts, however, results show that their ability is low also to simulate melt amounts at sites where sublimation represents larger percentages of total ablation. In fact, we

  9. Role of BWR MK I secondary containments in severe accident mitigation

    International Nuclear Information System (INIS)

    Greene, S.R.

    1986-01-01

    The recent advent of detailed containment analysis codes such as CONTAIN and MELCOR has facilitated the development of the first large-scale, architectural-based BWR secondary containment models. During the past year ORNL has developed detailed, plant-specific models of the Browns Ferry and Peach Bottom secondary containments, and applied these models in a variety of studies designed to evaluate the role and effectiveness of BWR secondary containments in severe accident mitigation. The topology and basis for these models is discussed, together with some of the emerging insights from these studies

  10. Thermo-hydrodynamic and inductive modelling of a glass melt elaborated in cold inductive crucible

    International Nuclear Information System (INIS)

    Sauvage, E.

    2009-11-01

    Within the context of a search for a new vitrification process for nuclear wastes with a replacement of the presently used metallic pot by an inductive cold crucible, this research thesis deals with the numerical modelling of this technology. After having recalled the interest of nuclear waste vitrification, this report presents the new process based on the use of a cold crucible, describing principles and objectives of this method, and the characteristic physical phenomena associated with the flow and the thermodynamics of the glassy melt in such a crucible. It also recalls and comments the existing works on modelling. The main objective of this research is then to demonstrate the feasibility of 3D thermo-hydraulic and inductive simulations. He describes and analyses the glass physical properties (electrical properties, viscosity, thermal properties), the electromagnetic, hydrodynamic and thermal phenomena. He presents in detail the bubbling mixing modelling, reports 3D induction and fluid mechanical coupling calculations, and specific thermal investigations (radiating transfers, thermal limit conditions)

  11. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  12. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Ott, L.J.

    1997-01-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  13. MELT-IIIB: an updated version of the melt code

    International Nuclear Information System (INIS)

    Tabb, K.K.; Lewis, C.H.; O'Dell, L.D.; Padilla, A. Jr.; Smith, D.E.; Wilburn, N.P.

    1979-04-01

    The MELT series is a reactor modeling code designed to investigate a wide variety of hypothetical accident conditions, particularly the transient overpower sequence. MELT-IIIB is the latest in the series

  14. A holistic model for the role of the axial melt lens at fast-spreading mid-ocean ridges

    Science.gov (United States)

    MacLeod, C. J.; Loocke, M. P.; Lissenberg, J. C. J.

    2016-12-01

    Axial melt lenses (AML) are melt or crystal mush1 bodies located at the dyke-gabbro transition beneath intermediate- and fast-spreading mid-ocean ridges (MORs)2,3. Although it is generally thought that AMLs play a major role in the storage and differentiation of mid-ocean ridge basalts (MORB)1, the melt compositions within the AML and its role in the accretion of the lower crust are heavily debated4-6. Here we present the first comprehensive study of the AML horizon at a fast-spreading MOR (Hess Deep, equatorial Pacific Ocean). We show that plagioclase and pyroxene within the AML are much too evolved to be in equilibrium with MORB, with mean An (54.85) and Mg# (65.01) consistent with derivation from basaltic andesite to andesite melts (Mg# 43-26). We propose that, in between decadal eruptions, the AML is predominantly crystal mush and is fed by small volumes of evolved interstitial melts. Short-lived, focused injection of primitive melt leads to mixing of primitive melts with the extant highly fractionated melt, and triggers eruptions. This model reconciles the paradoxical compositional mismatch between the volcanic and plutonic records with the geophysical characteristics of the AML, the short residence times of Pacific MORB phenocrysts, and the incompatible trace element over-enrichments in MORB. 1Marjanović, M. et al., 2015. Distribution of melt along the East Pacific Rise from 9°30' to 10°N from an amplitude variation with angle of incidence (AVA) technique. Geophys. J. Int. 203. 2Detrick, R. S. et al., 1987. Multi-channel seismic imaging of a crustal magma chamber along the EPR. Nature 326. 3Sinton, J. M. & Detrick, R. S., 1992. Mid-ocean ridge magma chambers. J. Geophys. Res. 97. 4Coogan, L. A., Thompson, G. & MacLeod, C. J., 2002. A textural and geochemical investigation of high level gabbros from the Oman ophiolite: implications for the role of the axial magma chamber at fast-spreading ridges. Lithos 63. 5Pan, Y. & Batiza, R., 2002. Mid-ocean ridge magma

  15. Geothermal flux and basal melt rate in the Dome C region inferred from radar reflectivity and heat modelling

    Science.gov (United States)

    Passalacqua, Olivier; Ritz, Catherine; Parrenin, Frédéric; Urbini, Stefano; Frezzotti, Massimo

    2017-09-01

    Basal melt rate is the most important physical quantity to be evaluated when looking for an old-ice drilling site, and it depends to a great extent on the geothermal flux (GF), which is poorly known under the East Antarctic ice sheet. Given that wet bedrock has higher reflectivity than dry bedrock, the wetness of the ice-bed interface can be assessed using radar echoes from the bedrock. But, since basal conditions depend on heat transfer forced by climate but lagged by the thick ice, the basal ice may currently be frozen whereas in the past it was generally melting. For that reason, the risk of bias between present and past conditions has to be evaluated. The objective of this study is to assess which locations in the Dome C area could have been protected from basal melting at any time in the past, which requires evaluating GF. We used an inverse approach to retrieve GF from radar-inferred distribution of wet and dry beds. A 1-D heat model is run over the last 800 ka to constrain the value of GF by assessing a critical ice thickness, i.e. the minimum ice thickness that would allow the present local distribution of basal melting. A regional map of the GF was then inferred over a 80 km × 130 km area, with a N-S gradient and with values ranging from 48 to 60 mW m-2. The forward model was then emulated by a polynomial function to compute a time-averaged value of the spatially variable basal melt rate over the region. Three main subregions appear to be free of basal melting, two because of a thin overlying ice and one, north of Dome C, because of a low GF.

  16. Geothermal flux and basal melt rate in the Dome C region inferred from radar reflectivity and heat modelling

    Directory of Open Access Journals (Sweden)

    O. Passalacqua

    2017-09-01

    Full Text Available Basal melt rate is the most important physical quantity to be evaluated when looking for an old-ice drilling site, and it depends to a great extent on the geothermal flux (GF, which is poorly known under the East Antarctic ice sheet. Given that wet bedrock has higher reflectivity than dry bedrock, the wetness of the ice–bed interface can be assessed using radar echoes from the bedrock. But, since basal conditions depend on heat transfer forced by climate but lagged by the thick ice, the basal ice may currently be frozen whereas in the past it was generally melting. For that reason, the risk of bias between present and past conditions has to be evaluated. The objective of this study is to assess which locations in the Dome C area could have been protected from basal melting at any time in the past, which requires evaluating GF. We used an inverse approach to retrieve GF from radar-inferred distribution of wet and dry beds. A 1-D heat model is run over the last 800 ka to constrain the value of GF by assessing a critical ice thickness, i.e. the minimum ice thickness that would allow the present local distribution of basal melting. A regional map of the GF was then inferred over a 80 km  ×  130 km area, with a N–S gradient and with values ranging from 48 to 60 mW m−2. The forward model was then emulated by a polynomial function to compute a time-averaged value of the spatially variable basal melt rate over the region. Three main subregions appear to be free of basal melting, two because of a thin overlying ice and one, north of Dome C, because of a low GF.

  17. Heat transfer in melt ponds with convection and radiative heating: observationally-inspired modelling

    Science.gov (United States)

    Wells, A.; Langton, T.; Rees Jones, D. W.; Moon, W.; Kim, J. H.; Wilkinson, J.

    2016-12-01

    Melt ponds have key impacts on the evolution of Arctic sea ice and summer ice melt. Small changes to the energy budget can have significant consequences, with a net heat-flux perturbation of only a few Watts per square metre sufficient to explain the thinning of sea ice over recent decades. Whilst parameterisations of melt-pond thermodynamics often assume that pond temperatures remain close to the freezing point, recent in-situ observations show more complex thermal structure with significant diurnal and synoptic variability. We here consider the energy budget of melt ponds and explore the role of internal convective heat transfer in determining the thermal structure within the pond in relatively calm conditions with low winds. We quantify the energy fluxes and temperature variability using two-dimensional direct numerical simulations of convective turbulence within a melt pond, driven by internal radiative heating and surface fluxes. Our results show that the convective flow dynamics are modulated by changes to the incoming radiative flux and sensible heat flux at the pond surface. The evolving pond surface temperature controls the outgoing longwave emissions from the pond. Hence the convective flow modifies the net energy balance of a melt pond, modulating the relative fractions of the incoming heat flux that is re-emitted to the atmosphere or transferred downward into the sea ice to drive melt.

  18. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  19. Water level measurement uncertainty during BWR instability

    International Nuclear Information System (INIS)

    Torok, R.C.; Derbidge, T.C.; Healzer, J.M.

    1994-01-01

    This paper addresses the performance of the water-level measurement system in a boiling water reactor (BWR) during severe instability oscillations which, under some circumstances, can occur during an anticipated transient without SCRAM (ATWS). Test data from a prototypical mock-up of the water-level measurement system was used to refine and calibrate a water-level measurement system model. The model was then used to predict level measurement system response, using as boundary conditions vessel pressures calculated by ppercase RETRAN for an ATWS/instability event.The results of the study indicate that rapid pressure changes in the reactor pressure vessel which cause oscillations in downcomer water level, coupled with differences in instrument line lengths, can produce errors in the sensed water level. Using nominal parameters for the measurement system components, a severe instability transient which produced a 0.2 m peak-to-minimum water-level oscillation in the vessel downcomer was predicted to produce pressure difference equivalent to a 0.7 m level oscillation at the input to the differential pressure transmitter, 0.5 m oscillation at the output of the transmitter, and an oscillation of 0.3 m on the water-level indicator in the control room. The level measurement system error, caused by downcomer water-level oscillations and instrument line length differential, is mitigated by damping both in the differential pressure transmitter used to infer level and in the control room display instrument. ((orig.))

  20. Dendritic solidification in undercooled Ni-Zr-Al melts: Experiments and modeling

    Energy Technology Data Exchange (ETDEWEB)

    Galenko, P.K., E-mail: Peter.Galenko@dlr.de [Institut fuer Materialsphysik im Weltraum, Deutsches Zentrum fuer Luft- und Raumfahrt (DLR), D-51170 Koeln (Germany); Reutzel, S.; Herlach, D.M. [Institut fuer Materialsphysik im Weltraum, Deutsches Zentrum fuer Luft- und Raumfahrt (DLR), D-51170 Koeln (Germany); Fries, S.G. [ICAMS, Ruhr-Universitaet Bochum, Stiepeler Strasse 129, D-44780 Bochum (Germany)] [SGF Scientific Consultancy, Arndtstr 9, D-52064 Aachen (Germany); Steinbach, I. [ICAMS, Ruhr-Universitaet Bochum, Stiepeler Strasse 129, D-44780 Bochum (Germany); Apel, M. [ACCESS eV, Intzestrasse 5, D-52072 Aachen (Germany)

    2009-12-15

    The kinetics of dendritic solidification in a ternary Ni{sub 98}Zr{sub 1}Al{sub 1} alloy is investigated experimentally in a range of melt undercoolings 40K{<=}{Delta}T{<=}320K. The growth velocity is measured for samples processed by the electromagnetic levitation technique using a high-speed video camera. With {Delta}T{<=}220K the measured growth rates are the same as those of a binary Ni{sub 99}Zr{sub 1} alloy. In the regime of rapid solidification, especially within the regime of thermal dendritic growth at {Delta}T{>=}220K, growth rates are decreased. Sharp-interface modeling predicts growth rates over the whole range of undercooling. Phase-field simulations give quantitative predictions for the dendritic growth velocity in the solute-controlled growth regime. Results show that the composition and temperature dependency of the thermodynamic data, e.g. liquidus slope and solute partition coefficient, are important for describing the alloys. Our findings give improved sharp-interface model predictions compared to calculations based on an approximation of the thermodynamic data derived from binary phase diagrams.

  1. The impact of BWR MK I primary containment failure dynamics on secondary containment integrity

    International Nuclear Information System (INIS)

    Greene, S.R.

    1987-01-01

    During the past four years, the ORNL BWRSAT Program has developed a series of increasingly sophisticated BWR secondary containment models. These models have been applied in a variety of studies to evaluate the severe accident mitigation capability of BWR secondary containments. This paper describes the results of a recent ORNL study of the impact of BWR MK I primary containment failure dynamics on secondary containment integrity. A 26-cell MELCOR Browns Ferry secondary containment model is described and the predicted thermodynamic response of the secondary containment to a variety of postulated primary containment failure modes is presented. The effects of primary containment failure location, timing, and ultimate hole size on secondary containment response is investigated, and the potential impact of hydrogen deflagrations on secondary containment integrity is explored

  2. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  3. Development of Physics-Based Numerical Models for Uncertainty Quantification of Selective Laser Melting Processes

    Data.gov (United States)

    National Aeronautics and Space Administration — The goal of the proposed research is to characterize the influence of process parameter variability inherent to Selective Laser Melting (SLM) and performance effect...

  4. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  5. Development and validation of a new LBM-MRT hybrid model with enthalpy formulation for melting with natural convection

    Energy Technology Data Exchange (ETDEWEB)

    Miranda Fuentes, Johann [Université de Lyon, CNRS, UMR5008, F-69622 Villeurbanne (France); INSA-Lyon, CETHIL, F-69621 Villeurbanne (France); Kuznik, Frédéric, E-mail: frederic.kuznik@insa-lyon.fr [Université de Lyon, CNRS, UMR5008, F-69622 Villeurbanne (France); INSA-Lyon, CETHIL, F-69621 Villeurbanne (France); Johannes, Kévyn; Virgone, Joseph [Université de Lyon, CNRS, UMR5008, F-69622 Villeurbanne (France); Université Lyon 1, CETHIL, F-69622 Villeurbanne (France)

    2014-01-17

    This article presents a new model to simulate melting with natural convection of a phase change material. For the phase change problem, the enthalpy formulation is used. Energy equation is solved by a finite difference method, whereas the fluid flow is solved by the multiple relaxation time (MRT) lattice Boltzmann method. The model is first verified and validated using the data from the literature. Then, the model is applied to a tall brick filled with a fatty acid eutectic mixture and the results are presented. The main results are (1) the spatial convergence rate is of second order, (2) the new model is validated against data from the literature and (3) the natural convection plays an important role in the melting process of the fatty acid mixture considered in our work.

  6. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  7. Recent technology for BWR operator training simulators

    International Nuclear Information System (INIS)

    Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

    1990-01-01

    As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)

  8. Multi-physics modeling of single/multiple-track defect mechanisms in electron beam selective melting

    International Nuclear Information System (INIS)

    Yan, Wentao; Ge, Wenjun; Qian, Ya; Lin, Stephen; Zhou, Bin; Liu, Wing Kam; Lin, Feng; Wagner, Gregory J.

    2017-01-01

    Metallic powder bed-based additive manufacturing technologies have many promising attributes. The single track acts as one fundamental building unit, which largely influences the final product quality such as the surface roughness and dimensional accuracy. A high-fidelity powder-scale model is developed to predict the detailed formation processes of single/multiple-track defects, including the balling effect, single track nonuniformity and inter-track voids. These processes are difficult to observe in experiments; previous studies have proposed different or even conflicting explanations. Our study clarifies the underlying formation mechanisms, reveals the influence of key factors, and guides the improvement of fabrication quality of single tracks. Additionally, the manufacturing processes of multiple tracks along S/Z-shaped scan paths with various hatching distance are simulated to further understand the defects in complex structures. The simulations demonstrate that the hatching distance should be no larger than the width of the remelted region within the substrate rather than the width of the melted region within the powder layer. Thus, single track simulations can provide valuable insight for complex structures.

  9. GEOMETRIC COMPLEXITY ANALYSIS IN AN INTEGRATIVE TECHNOLOGY EVALUATION MODEL (ITEM FOR SELECTIVE LASER MELTING (SLM#

    Directory of Open Access Journals (Sweden)

    S. Merkt

    2012-01-01

    Full Text Available

    ENGLISH ABSTRACT: Selective laser melting (SLM is becoming an economically viable choice for manufacturing complex serial parts. This paper focuses on a geometric complexity analysis as part of the integrative technology evaluation model (ITEM presented here. In contrast to conventional evaluation methodologies, the ITEM considers interactions between product and process innovations generated by SLM. The evaluation of manufacturing processes that compete with SLM is the main goal of ITEM. The paper includes a complexity analysis of a test part from Festo AG. The paper closes with a discussion of how the expanded design freedom of SLM can be used to improve company operations, and how the complexity analysis presented here can be seen as a starting point for feature-based complexity analysis..

    AFRIKAANSE OPSOMMING: Selektiewe lasersmelting word geleidelik ’n gangbare ekonomiese keuse vir die vervaar-diging van opeenvolgende komplekse onderdele. Die navorsing is toegespits op die ontleding van meetkundige kompleksiteit as ’n gedeelte van ’n integrerende tegnologiese evalueringsmodel. Gemeet teen konvensionele evalueringsmodelle behandel die genoemde metode interaksies tussen produkte- en prosesinnovasies wat gegenereer word. Die navorsing behandel ’n kompleksiteitsontleding van ’n toetsonderdeel van die firma FESTO AG. Die resultaat toon hoe kompleksiteits-analise gebruik kan word as die vertrekpunt vir eienskapsgebaseerde analise.

  10. BWR full integral simulation test (FIST) pretest predictions with TRACBO2

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.

    1984-01-01

    The Full Integral Simulation Test program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An analytical method development program is underway to extend the BWR-TRAC computer code to model reactor kinetics and major interfacing systems, including balance-of-plant, to improve application modeling flexibility, and to reduce computer running time. An experimental program is underway in a new single bundle system test facility to extend the large break loss-of-coolant accident LOCA data base to small breaks and operational transients. And a method qualification program is underway to test TRACBO2 against experiments in the FIST facility. The recently completed Phase 1 period included a series of LOCA and power transient tests, and successful pretest analysis of the large and small break LOCA tests with TRACBO2. These comparisons demonstrate BWR-TRAC capability for small and large break analysis, and provide detailed understanding of the phenomena

  11. A thermodynamic model for predicting surface melting and overheating of different crystal planes in BCC, FCC and HCP pure metallic thin films

    International Nuclear Information System (INIS)

    Jahangir, Vafa; Riahifar, Reza; Sahba Yaghmaee, Maziar

    2016-01-01

    In order to predict as well as study the surface melting phenomena in contradiction to surface overheating, a generalized thermodynamics model including the surface free energy of solid and the melt state along with the interfacial energy of solid–liquid (melt on substrate) has been introduced. In addition, the effect of different crystal structures of surfaces in fcc, bcc and hcp metals was included in surface energies as well as in the atomistic model. These considerations lead us to predict surface melting and overheating as two contradictory melting phenomena. The results of the calculation are demonstrated on the example of Pb and Al thin films in three groups of (100), (110) and (111) surface planes. Our conclusions show good agreement with experimental results and other theoretical investigations. Moreover, a computational algorithm has been developed which enables users to investigate the surface melt or overheating of single component metallic thin film with variable crystal structures and different crystalline planes. This model and developed software can be used for studying all related surface phenomena. - Highlights: • Investigating the surface melting and overheating phenomena • Effect of crystal orientations, surface energies, geometry and different atomic surface layers • Developing a computational algorithm and its related code (free-software SMSO-Ver1) • Thickness and orientation of surface plane dominate the surface melting or overheating. • Total excess surface energy as a function of thickness and temperature explains melting.

  12. A thermodynamic model for predicting surface melting and overheating of different crystal planes in BCC, FCC and HCP pure metallic thin films

    Energy Technology Data Exchange (ETDEWEB)

    Jahangir, Vafa, E-mail: vafa.jahangir@yahoo.com; Riahifar, Reza, E-mail: reza_rfr@yahoo.com; Sahba Yaghmaee, Maziar, E-mail: fkmsahba@uni-miskolc.hu

    2016-03-31

    In order to predict as well as study the surface melting phenomena in contradiction to surface overheating, a generalized thermodynamics model including the surface free energy of solid and the melt state along with the interfacial energy of solid–liquid (melt on substrate) has been introduced. In addition, the effect of different crystal structures of surfaces in fcc, bcc and hcp metals was included in surface energies as well as in the atomistic model. These considerations lead us to predict surface melting and overheating as two contradictory melting phenomena. The results of the calculation are demonstrated on the example of Pb and Al thin films in three groups of (100), (110) and (111) surface planes. Our conclusions show good agreement with experimental results and other theoretical investigations. Moreover, a computational algorithm has been developed which enables users to investigate the surface melt or overheating of single component metallic thin film with variable crystal structures and different crystalline planes. This model and developed software can be used for studying all related surface phenomena. - Highlights: • Investigating the surface melting and overheating phenomena • Effect of crystal orientations, surface energies, geometry and different atomic surface layers • Developing a computational algorithm and its related code (free-software SMSO-Ver1) • Thickness and orientation of surface plane dominate the surface melting or overheating. • Total excess surface energy as a function of thickness and temperature explains melting.

  13. ZZ BWRSB-RINGHALS1, Stability Benchmark Data from BWR RINGHALS-1

    International Nuclear Information System (INIS)

    2002-01-01

    Description of program or function: The purpose of this benchmark is to enable code developers to test their codes and also to validate the predictive capability of their respective codes and models for BWR stability analysis. Emphasis is put on the modelling of flow dynamics of the reactor core and in-vessel flow loop wit detailed neutronic and thermodynamic feedback. The secondary systems as well as the control and production systems will be neglected. Data provided comes from measurements in beginning of cycle (BOC) 14, 15, 16 and 17 and middle of cycle (MOC) 16 in the Swedish BWR reactor Ringhals 1. For these measurements complete data sets are given

  14. Complete BWR--EM LOCA analysis using the WRAP--EM system

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

    1979-01-01

    The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

  15. Mixed oxide thermal behaviour at BOL: COMETHE III-J models and impact on power-to-melt

    International Nuclear Information System (INIS)

    Vliet, J. van

    1979-01-01

    The mixed oxide thermal behaviour at beginning of life is very important because it can impose a limitation to the fuel pin peak power, and therefore to the reactor thermal output. The relevant physical processes leading to fuel restructuring are modelled in COMETHE III-J in a kinetic way. This ensures that the temperature and power history are properly taken into account. These models are described and their impact on the calculated power to melt early in life is analysed. (author)

  16. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    1 - Description of program or function: THALES, which stands for 'Thermal Hydraulic Analysis of Loss-of-coolant, Emergency core cooling and Severe core damage', is a computer code system for analyzing progression of core melt accident of light water reactors. The code was developed for Level 2 PSA (probabilistic safety assessment) and applicable to a wide range of postulated accident scenarios. Its outcomes are thermal hydraulic conditions in the reactor coolant system and the containment which are necessary for analyzing fission product release and transport behavior during the accident. The code system consists of following three member codes: (1) THALES-PM for accident progression in the primary and the secondary system of PWRs, (2) THALES-BM for accident progression in the reactor coolant system of BWRs, and (3) THALES-CV for accident progression in the containment of PWRs and BWRs. The THALES-PM and the THALES-BM codes carry out two categories of analysis. The first one is overall thermal-hydraulic analysis in the reactor coolant system. The reactor coolant system is divided into multi-volumes and each volume is further separated into a liquid region and a gas region by a movable mixture level. System pressure, mixture level in each volume, coolant temperature in each region, flow rate between volumes, etc. are calculated. The other one is core heatup and meltdown analysis. The reactor core is radially and axially divided into many nodes. Fuel and cladding temperature, cladding oxidation rate, hydrogen generation rate, core melt fraction, etc. are calculated. The THALES-CV code is for containment response analysis. It divides the containment into multiple compartments, each of which is further separated into a liquid region and a gas region by a movable mixture level. Containment pressure, mixture level in each compartment, coolant temperature in each region, flow rate between compartments, etc. are calculated. The code can treat coolant blowdown from the

  17. A thermodynamical model for the surface tension of silicate melts in contact with H2O gas

    Science.gov (United States)

    Colucci, Simone; Battaglia, Maurizio; Trigila, Raffaello

    2016-01-01

    Surface tension plays an important role in the nucleation of H2O gas bubbles in magmatic melts and in the time-dependent rheology of bubble-bearing magmas. Despite several experimental studies, a physics based model of the surface tension of magmatic melts in contact with H2O is lacking. This paper employs gradient theory to develop a thermodynamical model of equilibrium surface tension of silicate melts in contact with H2O gas at low to moderate pressures. In the last decades, this approach has been successfully applied in studies of industrial mixtures but never to magmatic systems. We calibrate and verify the model against literature experimental data, obtained by the pendant drop method, and by inverting bubble nucleation experiments using the Classical Nucleation Theory (CNT). Our model reproduces the systematic decrease in surface tension with increased H2O pressure observed in the experiments. On the other hand, the effect of temperature is confirmed by the experiments only at high pressure. At atmospheric pressure, the model shows a decrease of surface tension with temperature. This is in contrast with a number of experimental observations and could be related to microstructural effects that cannot be reproduced by our model. Finally, our analysis indicates that the surface tension measured inverting the CNT may be lower than the value measured by the pendant drop method, most likely because of changes in surface tension controlled by the supersaturation.

  18. BWR Refill-Reflood Program. Final report

    International Nuclear Information System (INIS)

    Myers, L.L.

    1983-09-01

    The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

  19. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1983-01-01

    A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  20. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  1. Modelling the influence of the gas to melt ratio on the fraction solid of the surface in spray formed billets

    DEFF Research Database (Denmark)

    Hattel, Jesper Henri; Pryds, Nini

    2006-01-01

    the atomisation stage taking thermal coupling into consideration and the deposition of the droplets at the surface of the billet taking geometrical aspects such as shading into account. The coupling between these two models is accomplished by ensuring that the total droplet size distribution of the spray......In this paper, the relationship between the Gas to Melt Ratio (GMR) and the solid fraction of an evolving billet surface is investigated numerically. The basis for the analysis is a recently developed integrated procedure for modelling the entire spray forming process. This model includes...... is the summation of “local” droplet size distributions along the r-axis of the spray cone. The criterion for a successful process has been a predefined process window characterised by a desired solid fraction range at a certain distance from the atomizer. Inside this process window, the gas and melt flows have...

  2. BWR radiation exposure--experience and projection

    International Nuclear Information System (INIS)

    Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

    1979-01-01

    The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

  3. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

  4. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  5. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  6. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  7. Melt analysis of mismatch amplification mutation assays (Melt-MAMA: a functional study of a cost-effective SNP genotyping assay in bacterial models.

    Directory of Open Access Journals (Sweden)

    Dawn N Birdsell

    Full Text Available Single nucleotide polymorphisms (SNPs are abundant in genomes of all species and biologically informative markers extensively used across broad scientific disciplines. Newly identified SNP markers are publicly available at an ever-increasing rate due to advancements in sequencing technologies. Efficient, cost-effective SNP genotyping methods to screen sample populations are in great demand in well-equipped laboratories, but also in developing world situations. Dual Probe TaqMan assays are robust but can be cost-prohibitive and require specialized equipment. The Mismatch Amplification Mutation Assay, coupled with melt analysis (Melt-MAMA, is flexible, efficient and cost-effective. However, Melt-MAMA traditionally suffers from high rates of assay design failures and knowledge gaps on assay robustness and sensitivity. In this study, we identified strategies that improved the success of Melt-MAMA. We examined the performance of 185 Melt-MAMAs across eight different pathogens using various optimization parameters. We evaluated the effects of genome size and %GC content on assay development. When used collectively, specific strategies markedly improved the rate of successful assays at the first design attempt from ~50% to ~80%. We observed that Melt-MAMA accurately genotypes across a broad DNA range (~100 ng to ~0.1 pg. Genomic size and %GC content influence the rate of successful assay design in an independent manner. Finally, we demonstrated the versatility of these assays by the creation of a duplex Melt-MAMA real-time PCR (two SNPs and conversion to a size-based genotyping system, which uses agarose gel electrophoresis. Melt-MAMA is comparable to Dual Probe TaqMan assays in terms of design success rate and accuracy. Although sensitivity is less robust than Dual Probe TaqMan assays, Melt-MAMA is superior in terms of cost-effectiveness, speed of development and versatility. We detail the parameters most important for the successful application of

  8. Analysis of a BWR direct cycle forced circulation power plants operation

    International Nuclear Information System (INIS)

    Andrade, G.G. de.

    1973-01-01

    First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

  9. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  10. Subchannel analysis of a critical power test, using simulated BWR 8x8 fuel assembly

    International Nuclear Information System (INIS)

    Mitsutake, T.; Terasaka, H.; Yoshimura, K.; Oishi, M.; Inoue, A.; Akiyama, M.

    1990-01-01

    Critical power predictions have been compared with the critical power test data obtained in simulated BWR 8x8 fuel rod assemblies. Two analytical methods for the critical power prediction in rod assemblies are used in the prediction, which are the subchannel analysis using the COBRA/BWR subchannel computer code with empirical critical heat flux (CHF) correlations and the liquid film dryout estimation using the CRIPP-3F 'multi-fluid' computer code. Improvements in both the analytical methods were made for spacer effect modeling, though they were specific for application to the current BWR rod assembly type. In general a reasonable agreement was obtained, though comparisons, between the prediction and the obtained test data. (orig.)

  11. On the influence of debris in glacier melt modelling: a new temperature-index model accounting for the debris thickness feedback

    Science.gov (United States)

    Carenzo, Marco; Mabillard, Johan; Pellicciotti, Francesca; Reid, Tim; Brock, Ben; Burlando, Paolo

    2013-04-01

    The increase of rockfalls from the surrounding slopes and of englacial melt-out material has led to an increase of the debris cover extent on Alpine glaciers. In recent years, distributed debris energy-balance models have been developed to account for the melt rate enhancing/reduction due to a thin/thick debris layer, respectively. However, such models require a large amount of input data that are not often available, especially in remote mountain areas such as the Himalaya. Some of the input data such as wind or temperature are also of difficult extrapolation from station measurements. Due to their lower data requirement, empirical models have been used in glacier melt modelling. However, they generally simplify the debris effect by using a single melt-reduction factor which does not account for the influence of debris thickness on melt. In this paper, we present a new temperature-index model accounting for the debris thickness feedback in the computation of melt rates at the debris-ice interface. The empirical parameters (temperature factor, shortwave radiation factor, and lag factor accounting for the energy transfer through the debris layer) are optimized at the point scale for several debris thicknesses against melt rates simulated by a physically-based debris energy balance model. The latter has been validated against ablation stake readings and surface temperature measurements. Each parameter is then related to a plausible set of debris thickness values to provide a general and transferable parameterization. The new model is developed on Miage Glacier, Italy, a debris cover glacier in which the ablation area is mantled in near-continuous layer of rock. Subsequently, its transferability is tested on Haut Glacier d'Arolla, Switzerland, where debris is thinner and its extension has been seen to expand in the last decades. The results show that the performance of the new debris temperature-index model (DETI) in simulating the glacier melt rate at the point scale

  12. GLASS MELTING PHENOMENA, THEIR ORDERING AND MELTING SPACE UTILISATION

    Directory of Open Access Journals (Sweden)

    Němec L.

    2013-12-01

    Full Text Available Four aspects of effective glass melting have been defined – namely the fast kinetics of partial melting phenomena, a consideration of the melting phenomena ordering, high utilisation of the melting space, and effective utilisation of the supplied energy. The relations were defined for the specific melting performance and specific energy consumption of the glass melting process which involve the four mentioned aspects of the process and indicate the potentials of effective melting. The quantity “space utilisation” has been treated in more detail as an aspect not considered in practice till this time. The space utilisation was quantitatively defined and its values have been determined for the industrial melting facility by mathematical modelling. The definitions of the specific melting performance and specific energy consumption have been used for assessment of the potential impact of a controlled melt flow and high space utilisation on the melting process efficiency on the industrial scale. The results have shown that even the partial control of the melt flow, leading to the partial increase of the space utilisation, may considerably increase the melting performance, whereas a decrease of the specific energy consumption was determined to be between 10 - 15 %.

  13. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  14. General Electric Company analytical model for loss-of-coolant analysis in accordance with 10CFR50 appendix K, amendment No. 3: effect of steam environment on BWR core spray distribution

    International Nuclear Information System (INIS)

    1977-04-01

    The core spray sparger designs of the BWR/2 through BWR/5 product lines were verified by means of full-scale mock-ups tested in air at various flow conditions. In 1974, an overseas technical partner of General Electric reported that a steam environment changed the individual core spray nozzle patterns when compared to patterns measured in air. This document describes preliminary findings of how a steam environment alters the core spray nozzle pattern, and the actions which General Electric is pursuing to quantify the steam effects

  15. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    Hsiang-Shou Cheng; Diamond, D.J.

    1978-01-01

    The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

  16. RETRAN experience with BWR transients at Yankee Atomic Electric Company

    International Nuclear Information System (INIS)

    Ansari, A.A.F.; Cronin, J.T.; Slifer, B.C.

    1981-01-01

    Yankee Atomic Electric Company is actively involved in the development of licensing methods for BWR's. The computer code chosen for analyzing system response under transient conditions is RETRAN. This paper describes the RETRAN model developed for Vermont Yankee, and the results of the RETRAN checkout and qualification that has been achieved at YAEC through comparison of RETRAN predictions to the startup test results performed at the plant as part of the 100% power startup test program. In addition, abnormal operational transients typically analyzed for licensing are also presented

  17. Time-dependent coolant velocity measurements in an operating BWR

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.; Crowe, R.D.

    1980-01-01

    A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

  18. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

  19. Dynamic fragmentation of laser shock-melted tin: experiment and modelling

    Energy Technology Data Exchange (ETDEWEB)

    De Resseguier, T. [CNRS ENSMA, Lab Combust and Deton, F-86961 Futuroscope (France); Signor, L.; Dragon, A. [CNRS ENSMA, Mecan and Phys Mat Lab, F-86961 Futuroscope (France); Signor, L.; Roy, G. [CEA Valduc, 21 - Is-sur-Tille (France)

    2010-07-01

    Dynamic fragmentation of shock-loaded metals is an issue of considerable importance for both basic science and a variety of technological applications, such as pyrotechnics or inertial confinement fusion, the latter involving high energy laser irradiation of thin metallic shells. Whereas spall fracture in solid materials has been extensively studied for many years, little data can be found yet about the evolution of this phenomenon after partial or full melting on compression or on release. Here, we present an investigation of dynamic fragmentation in laser shock-melted tin, from the 'micro-spall' process (ejection of a cloud of fine droplets) occurring upon reflection of the compressive pulse from the target free surface, to the late rupture observed in the un-spalled melted layer (leading to the formation of larger spherical fragments). Experimental results consist of time-resolved velocity measurements and post-shock observations of recovered targets and fragments. They provide original information regarding the loss of tensile strength associated with melting, the cavitation mechanism likely to occur in the melted metal, the sizes of the subsequent fragments and their ejection velocities. A theoretical description based on an energetic approach adapted to the case of a liquid metal is implemented as a failure criterion in a one-dimensional hydro-code including a multi-phase equation of state for tin. The resulting predictions of the micro-spall process are compared with experimental data. In particular, the use of a new experimental technique to quantify the fragment size distributions leads to a much better agreement with theory than previously reported. Finally, a complementary approach focused on cavitation is proposed to evaluate the role of this phenomenon in the fragmentation of the melted metal. (authors)

  20. Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Aounallah, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2001-07-01

    This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

  1. Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

    International Nuclear Information System (INIS)

    Aounallah, Y.

    2001-01-01

    This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

  2. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  3. Numerical modeling of transferred arc melting bath heating; Modelisation numerique du chauffage de bains par arc transfere

    Energy Technology Data Exchange (ETDEWEB)

    Bouvier, A. [Electricite de France, 77 - Moret sur Loing (France). Direction des Etudes et Recherches; Trenty, L.; Guillot, J.B. [Ecole Centrale de Paris, Laboratoire EM2C. CNRS, 92 - Chatenay-Malabry (France); Delalondre, C. [Electricite de France (EDF), 78 - Chatou (France). Direction des Etudes et Recherches

    1997-12-31

    This paper presents the modeling of a transferred electric arc inside a bath of melted metal. After a recall of the context of the study, the problem of the modeling, which involves magnetohydrodynamic coupling inside the arc and the bath, is described. The equations that govern the phenomena inside the arc and the bath are recalled and the approach used for the modeling of the anode region of the arc is explained using a 1-D sub-model. The conditions of connection between arc and bath calculations are explained and calculation results obtained with a 200 kW laboratory furnace geometry are presented. (J.S.) 8 refs.

  4. Migrating Toward Fully 4-D Geodynamical Models of Asthenospheric Circulation and Melt Production at Mid-Ocean Ridges

    Science.gov (United States)

    van Dam, L.; Kincaid, C. R.; Pockalny, R. A.; Sylvia, R. T.; Hall, P. S.

    2017-12-01

    Lateral migration of mid-ocean ridge spreading centers is a well-documented phenomenon leading to asymmetric melt production and the surficial expressions thereof. This form of plate motion has been difficult to incorporate into both numerical and analogue geodynamical models, and consequently, current estimates of time-dependent flow, material transport, and melting in the mantle beneath ridges are lacking. To address this, we have designed and built an innovative research apparatus that allows for precise and repeatable simulations of mid-ocean ridge spreading and migration. Three pairs of counter-rotating belts with adjustable lateral orientations are scaled to simulate spreading at, and flow beneath, three 600km wide ridge segments with up to 300km transform offsets. This apparatus is attached to a drive system that allows us to test a full range of axis-parallel to axis-normal migration directions, and is suspended above a reservoir of viscous glucose syrup, a scaled analogue for the upper mantle, and neutrally buoyant tracers. We image plate-driven flow in the syrup with high-resolution digital cameras and use particle image velocimetry methods to obtain information about transport pathlines and flow-induced anisotropy. Suites of experiments are run with and without ridge migration to determine the overall significance of migration on spatial and temporal characteristics of shallow mantle flow. Our experiments cover an expansive parameter space by including various spreading rates, migration speeds and directions, degrees of spreading asymmetry, transform-offset lengths, and upper mantle viscosity conditions. Preliminary results highlight the importance of modeling migratory plate forces. Mantle material exhibits a significant degree of lateral transport, particularly between ridge segments and towards the melt triangle. Magma supply to the melting region is highly complex; parcels of material do not necessarily move along fixed streamlines, rather, they can

  5. Particle melting and particle/plasma interactions in DC and RF plasmas: a modeling study. (Volumes I and II)

    International Nuclear Information System (INIS)

    Wei, D.Y.C.

    1987-01-01

    Integral process models were developed to predict particle melting in both DC and RF plasmas. Specifically, a numerical model has been developed to predict the temperature history of particles injected in a low pressure DC plasma jet. The temperature and velocity fields of the plasma jet are predicted as a free jet by solving the parabolized Navier-Stokes equations using a spatial marching scheme. Correction factors were introduced to take into account non continuum effects encountered in the low pressure environment. The plasma jet profiles as well as the particle/plasma interactions under different jet pressure ratios (from underexpanded to overexpanded) were investigated. The flow and temperature fields in the RF plasma torch are calculated using the axisymmetric Navier-Stokes equations based on the primitive variables, along with pseudo two-dimensional electromagnetic field equations. Particle trajectories and heat transfer characteristics in both DC and RF plasmas are calculated using predicted plasma jet profiles. Particle melting efficiencies in both DC and RF plasmas are evaluated and compared using model alloy systems. Based on the theoretical considerations, an alternative route of plasma spraying process (hybrid plasma spraying process) is proposed. An evaluation of particle melting in hybrid plasma jets had indicated that further improvement in deposit properties could be made

  6. CoCrMo cellular structures made by Electron Beam Melting studied by local tomography and finite element modelling

    Energy Technology Data Exchange (ETDEWEB)

    Petit, Clémence [INSA de Lyon, MATEIS CNRS UMR5510, Université de Lyon, 69621 Villeurbanne (France); Maire, Eric, E-mail: eric.maire@insa-lyon.fr [INSA de Lyon, MATEIS CNRS UMR5510, Université de Lyon, 69621 Villeurbanne (France); Meille, Sylvain; Adrien, Jérôme [INSA de Lyon, MATEIS CNRS UMR5510, Université de Lyon, 69621 Villeurbanne (France); Kurosu, Shingo; Chiba, Akihiko [Institute for Materials Research, Tohoku University, Sendai 980-0812 (Japan)

    2016-06-15

    The work focuses on the structural and mechanical characterization of Co-Cr-Mo cellular samples with cubic pore structure made by Electron Beam Melting (EBM). X-ray tomography was used to characterize the architecture of the sample. High resolution images were also obtained thanks to local tomography in which the specimen is placed close to the X-ray source. These images enabled to observe some defects due to the fabrication process: small pores in the solid phase, partially melted particles attached to the surface. Then, in situ compression tests were performed in the tomograph. The images of the deformed sample show a progressive buckling of the vertical struts leading to final fracture. The deformation initiated where the defects were present in the strut i.e. in regions with reduced local thickness. The finite element modelling confirmed the high stress concentrations of these weak points leading to the fracture of the sample. - Highlights: • CoCrMo samples fabricated by Electron Beam Melting (EBM) process are considered. • X-ray Computed Tomography is used to observe the structure of the sample. • The mechanical properties are tested thanks to an in situ test in the tomograph. • A finite element model is developed to model the mechanical behaviour.

  7. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  8. Moderator temperature coefficient in BWR core

    International Nuclear Information System (INIS)

    Naito, Yoshitaka

    1977-01-01

    Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

  9. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  10. Delivering high performance BWR fuel reliably

    International Nuclear Information System (INIS)

    Schardt, J.F.

    1998-01-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  11. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  12. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  13. Modelling of the isothermal replication of surface microstructures in polymer melts

    DEFF Research Database (Denmark)

    Rasmussen, Henrik Koblitz; Eriksson, Torbjörn Gerhard

    2005-01-01

    boundary condition. This allows an investigation of the effect of the rheological properties of the polymer melt on the ability of the material to fill small structures in a mould surface. Series of isothermal compression moulding experiments were performed with a polycarbonate (PC) and a polystyrene (PS...

  14. Point, surface and volumetric heat sources in the thermal modelling of selective laser melting

    NARCIS (Netherlands)

    Yang, Y.; Ayas, C.; Brabazon, Dermot; Naher, Sumsun; Ul Ahad, Inam

    2017-01-01

    Selective laser melting (SLM) is a powder based additive manufacturing technique suitable for producing high precision metal parts. However, distortions and residual stresses within products arise during SLM because of the high temperature gradients created by the laser heating. Residual stresses

  15. A semi-analytical thermal modelling approach for selective laser melting

    NARCIS (Netherlands)

    Yang, Y.; van Keulen, A.; Ayas, C.

    2018-01-01

    Selective laser melting (SLM) wherein a metal part is built in a layer-by-layer manner in a powder bed is a promising and versatile way for manufacturing components with complex geometry. However, components built by SLM suffer from substantial deformation of the part and residual stresses.

  16. Role of melt behavior in modifying oxidation distribution using an interface incorporated model in selective laser melting of aluminum-based material

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Dongdong, E-mail: dongdonggu@nuaa.edu.cn; Dai, Donghua [College of Materials Science and Technology, Nanjing University of Aeronautics and Astronautics, Yudao Street 29, Nanjing 210016 (China); Institute of Additive Manufacturing (3D Printing), Nanjing University of Aeronautics and Astronautics, Yudao Street 29, Nanjing 210016 (China)

    2016-08-28

    A transient three dimensional model for describing the molten pool dynamics and the response of oxidation film evolution in the selective laser melting of aluminum-based material is proposed. The physical difference in both sides of the scan track, powder-solid transformation and temperature dependent physical properties are taken into account. It shows that the heat energy tends to accumulate in the powder material rather than in the as-fabricated part, leading to the formation of the asymmetrical patterns of the temperature contour and the attendant larger dimensions of the molten pool in the powder phase. As a higher volumetric energy density is applied (≥1300 J/mm{sup 3}), a severe evaporation is produced with the upward direction of velocity vector in the irradiated powder region while a restricted operating temperature is obtained in the as-fabricated part. The velocity vector continuously changes from upward direction to the downward one as the scan speed increases from 100 mm/s to 300 mm/s, promoting the generation of the debris of the oxidation films and the resultant homogeneous distribution state in the matrix. For the applied hatch spacing of 50 μm, a restricted remelting phenomenon of the as-fabricated part is produced with the upward direction of the convection flow, significantly reducing the turbulence of the thermal-capillary convection on the breaking of the oxidation films, and therefore, the connected oxidation films through the neighboring layers are typically formed. The morphology and distribution of the oxidation are experimentally acquired, which are in a good agreement with the results predicted by simulation.

  17. Coordinated Hard Sphere Mixture (CHaSM): A fast approximate model for oxide and silicate melts at extreme conditions

    Science.gov (United States)

    Wolf, A. S.; Asimow, P. D.; Stevenson, D. J.

    2015-12-01

    Recent first-principles calculations (e.g. Stixrude, 2009; de Koker, 2013), shock-wave experiments (Mosenfelder, 2009), and diamond-anvil cell investigations (Sanloup, 2013) indicate that silicate melts undergo complex structural evolution at high pressure. The observed increase in cation-coordination (e.g. Karki, 2006; 2007) induces higher compressibilities and lower adiabatic thermal gradients in melts as compared with their solid counterparts. These properties are crucial for understanding the evolution of impact-generated magma oceans, which are dominated by the poorly understood behavior of silicates at mantle pressures and temperatures (e.g. Stixrude et al. 2009). Probing these conditions is difficult for both theory and experiment, especially given the large compositional space (MgO-SiO2-FeO-Al2O3-etc). We develop a new model to understand and predict the behavior of oxide and silicate melts at extreme P-T conditions (Wolf et al., 2015). The Coordinated Hard Sphere Mixture (CHaSM) extends the Hard Sphere mixture model, accounting for the range of coordination states for each cation in the liquid. Using approximate analytic expressions for the hard sphere model, this fast statistical method compliments classical and first-principles methods, providing accurate thermodynamic and structural property predictions for melts. This framework is applied to the MgO system, where model parameters are trained on a collection of crystal polymorphs, producing realistic predictions of coordination evolution and the equation of state of MgO melt over a wide P-T range. Typical Mg-coordination numbers are predicted to evolve continuously from 5.25 (0 GPa) to 8.5 (250 GPa), comparing favorably with first-principles Molecular Dynamics (MD) simulations. We begin extending the model to a simplified mantle chemistry using empirical potentials (generally accurate over moderate pressure ranges, consuming classical MD calculations. This approach also sheds light on the universality

  18. TRAB - A transient analysis program for BWR. Part 2

    International Nuclear Information System (INIS)

    Raety, H.; Rajamaeki, M.

    1991-05-01

    TRAB is a transient analysis code for BWRs developed at the Technical Research Centre of Finland. It models the phenomena in the interior of the BWR pressure vessel and in related subsystems. The core model of TRAB can be used separately for LWR modelling. For PWR modelling the core model of TRAB is connected to circuit model SMABRE to form the SMATRA code. This report is a user's manual and documents the structure, contents and preparation of input for TRAB. The structure of TRAB input is very flexible, featuring input groups and subgroups identified with keywords and given in any order as well as data items in free format, freely mixed with explanatory texts. Users interface of the code can be used for modelling within input: through normal input it is possible to create new submodels. These may be functional or tabulated dependencies of the code variables, different types of delays, or ordinary linear differential equations

  19. EVALUATION OF GLACIER MELT CONTRIBUTION TO RUNOFF IN THE NORTH CAUCASUS ALPINE CATCHMENTS USING ISOTOPIC METHODS AND ENERGY BALANCE MODELING

    Directory of Open Access Journals (Sweden)

    E. Rets

    2017-01-01

    Full Text Available Frequency and intensity of river floods rise observed in the North Caucasus during last decades is considered to be driven by recent climate change. In order to predict possible future trends in extreme hydrological events in the context of climate change, it is essential to estimate the contribution of different feed sources in complicated flow-forming processes in the alpine part of the North Caucasus. A study was carried out for the Djankuat River basin, the representative for the North Caucasus system. Simultaneous measurements of electrical conductivity, isotopic and ion balance equations, and energy balance modeling of ice and snow melt were used to evaluate the contribution of different sources and processes in the Djankuat River runoff regime formation. A forecast of possible future changes in the Djankuat glacier melting regime according to the predicted climate changes was done.

  20. Continuum model of tensile fracture of metal melts and its application to a problem of high-current electron irradiation of metals

    International Nuclear Information System (INIS)

    Mayer, Alexander E.; Mayer, Polina N.

    2015-01-01

    A continuum model of the metal melt fracture is formulated on the basis of the continuum mechanics and theory of metastable liquid. A character of temperature and strain rate dependences of the tensile strength that is predicted by the continuum model is verified, and parameters of the model are fitted with the use of the results of the molecular dynamics simulations for ultra-high strain rates (≥1–10/ns). A comparison with experimental data from literature is also presented for Al and Ni melts. Using the continuum model, the dynamic tensile strength of initially uniform melts of Al, Cu, Ni, Fe, Ti, and Pb within a wide range of strain rates (from 1–10/ms to 100/ns) and temperatures (from melting temperature up to 70–80% of critical temperature) is calculated. The model is applied to numerical investigation of a problem of the high-current electron irradiation of Al, Cu, and Fe targets

  1. Analysis and evaluation of the ASTEC model basis on fission product and aerosol release phenomena from melts. 3. Technical report

    International Nuclear Information System (INIS)

    Agethen, K.; Koch, M.K.

    2016-04-01

    The present report is the 3 rd Technical Report within the research project ''ASMO'' founded by the German Federal Ministry for Economic Affairs and Energy (BMWi 1501433) and projected at the Chair of Energy Systems and Energy Economics (LEE) within the workgroup Reactor Simulation and Safety at the Ruhr-Universitaet Bochum (RUB). The focus in this report is set on the release of fission products and the contribution to the source term, which is formed in the late phase after failure of the reactor pressure vessel during MCCI. By comparing the RUB simulation results including the fission product release rates with further simulations of GRS and VEIKI it can be indicated that the simulations have a high sensitivity in respect to the melting point temperature. It can be noted that the release rates are underestimated for most fission product species with the current model. Especially semi-volatile fission products and the lanthanum release is underestimated by several orders of magnitude. Based on the ACE experiment L2, advanced considerations are presented concerning the melt temperature, the gas temperature, the segregation and a varied melt configuration. Furthermore, the influence of the gas velocity is investigated. This variation of the gas velocity causes an underestimation of the release rates compared to the RUB base calculation. A model extension to oxidic species for lanthanum and ruthenium shows a significant improvement of the simulation results. In addition, the MEDICIS module has been enhanced to document the currently existing species, are displayed in a *.ist-file. This expansion shows inconsistencies between the melt composition and the fission product composition. Based on these results, there are still some difficulties regarding the release of fission products in the MEDICIS module and the interaction with the material data base (MOB) which needs further investigation.

  2. Building optimal regression tree by ant colony system-genetic algorithm: Application to modeling of melting points

    Energy Technology Data Exchange (ETDEWEB)

    Hemmateenejad, Bahram, E-mail: hemmatb@sums.ac.ir [Department of Chemistry, Shiraz University, Shiraz (Iran, Islamic Republic of); Medicinal and Natural Products Chemistry Research Center, Shiraz University of Medical Sciences, Shiraz (Iran, Islamic Republic of); Shamsipur, Mojtaba [Department of Chemistry, Razi University, Kermanshah (Iran, Islamic Republic of); Zare-Shahabadi, Vali [Young Researchers Club, Mahshahr Branch, Islamic Azad University, Mahshahr (Iran, Islamic Republic of); Akhond, Morteza [Department of Chemistry, Shiraz University, Shiraz (Iran, Islamic Republic of)

    2011-10-17

    Highlights: {yields} Ant colony systems help to build optimum classification and regression trees. {yields} Using of genetic algorithm operators in ant colony systems resulted in more appropriate models. {yields} Variable selection in each terminal node of the tree gives promising results. {yields} CART-ACS-GA could model the melting point of organic materials with prediction errors lower than previous models. - Abstract: The classification and regression trees (CART) possess the advantage of being able to handle large data sets and yield readily interpretable models. A conventional method of building a regression tree is recursive partitioning, which results in a good but not optimal tree. Ant colony system (ACS), which is a meta-heuristic algorithm and derived from the observation of real ants, can be used to overcome this problem. The purpose of this study was to explore the use of CART and its combination with ACS for modeling of melting points of a large variety of chemical compounds. Genetic algorithm (GA) operators (e.g., cross averring and mutation operators) were combined with ACS algorithm to select the best solution model. In addition, at each terminal node of the resulted tree, variable selection was done by ACS-GA algorithm to build an appropriate partial least squares (PLS) model. To test the ability of the resulted tree, a set of approximately 4173 structures and their melting points were used (3000 compounds as training set and 1173 as validation set). Further, an external test set containing of 277 drugs was used to validate the prediction ability of the tree. Comparison of the results obtained from both trees showed that the tree constructed by ACS-GA algorithm performs better than that produced by recursive partitioning procedure.

  3. Boron concentration evolution in the temporary curtains of a BWR reactor. Burcur code

    International Nuclear Information System (INIS)

    Cano Aguado, M.; Perlado Martin, J.M.; Minguez Torres, E.

    1977-01-01

    The theoretical model and the user's guide of the code Burcur is included. This code analyzes the burnable poison concentration of the temporary curtains as a function of time, for BWR reactors of the 7 x 7 design. The computing time being reasonably short, the number of burnup steps is as high as necessary.(author) [es

  4. Disorder effect on heat capacity, self-diffusion coefficient, and choosing best potential model for melting temperature, in gold–copper bimetallic nanocluster with 55 atoms

    International Nuclear Information System (INIS)

    Taherkhani, Farid; Akbarzadeh, Hamed; Feyzi, Mostafa; Rafiee, Hamid Reza

    2015-01-01

    Molecular dynamics simulation has been implemented for doping effect on melting temperature, heat capacity, self-diffusion coefficient of gold–copper bimetallic nanostructure with 55 total gold and copper atom numbers and its bulk alloy. Trend of melting temperature for gold–copper bimetallic nanocluster is not same as melting temperature copper–gold bulk alloy. Molecular dynamics simulation of our result regarding bulk melting temperature is consistence with available experimental data. Molecular dynamics simulation shows that melting temperature of gold–copper bimetallic nanocluster increases with copper atom fraction. Semi-empirical potential model and quantum Sutton–Chen potential models do not change melting temperature trend with copper doping of gold–copper bimetallic nanocluster. Self-diffusion coefficient of copper atom is greater than gold atom in gold–copper bimetallic nanocluster. Semi-empirical potential within the tight-binding second moment approximation as new application potential model for melting temperature of gold–copper bulk structure shows better result in comparison with EAM, Sutton–Chen potential, and quantum Sutton–Chen potential models

  5. Disorder effect on heat capacity, self-diffusion coefficient, and choosing best potential model for melting temperature, in gold–copper bimetallic nanocluster with 55 atoms

    Energy Technology Data Exchange (ETDEWEB)

    Taherkhani, Farid, E-mail: faridtaherkhani@gmail.com, E-mail: f.taherkhani@razi.ac.ir [Razi University, Department of Physical Chemistry (Iran, Islamic Republic of); Akbarzadeh, Hamed [Hakim Sabzevari University, Department of Chemistry (Iran, Islamic Republic of); Feyzi, Mostafa; Rafiee, Hamid Reza [Razi University, Department of Physical Chemistry (Iran, Islamic Republic of)

    2015-01-15

    Molecular dynamics simulation has been implemented for doping effect on melting temperature, heat capacity, self-diffusion coefficient of gold–copper bimetallic nanostructure with 55 total gold and copper atom numbers and its bulk alloy. Trend of melting temperature for gold–copper bimetallic nanocluster is not same as melting temperature copper–gold bulk alloy. Molecular dynamics simulation of our result regarding bulk melting temperature is consistence with available experimental data. Molecular dynamics simulation shows that melting temperature of gold–copper bimetallic nanocluster increases with copper atom fraction. Semi-empirical potential model and quantum Sutton–Chen potential models do not change melting temperature trend with copper doping of gold–copper bimetallic nanocluster. Self-diffusion coefficient of copper atom is greater than gold atom in gold–copper bimetallic nanocluster. Semi-empirical potential within the tight-binding second moment approximation as new application potential model for melting temperature of gold–copper bulk structure shows better result in comparison with EAM, Sutton–Chen potential, and quantum Sutton–Chen potential models.

  6. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    International Nuclear Information System (INIS)

    Phung, Viet-Anh; Galushin, Sergey; Raub, Sebastian; Goronovski, Andrei; Villanueva, Walter; Kööp, Kaspar; Grishchenko, Dmitry; Kudinov, Pavel

    2016-01-01

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small ( 100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input

  7. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2016-08-15

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

  8. Comparative analysis of the simulation of the instantaneous closing of the discharge valve of a recirculation loop of a BWR with a model of recirculation loop with 2 jet pumps and another model with 20 jet pumps using RELAP5/SCDAPSIM Mod. 3.4

    International Nuclear Information System (INIS)

    Araiza M, E.; Ortiz V, J.; Martinez C, E.; Amador G, R.; Castillo D, R.

    2016-09-01

    This work presents the results of the simulation of the instantaneous closing of the water hammer, of a recirculation loop using two different arrangements in the loops. One of these arrangements corresponds to the traditional model that uses only two jet pumps to simulate the twenty pumps of the two recirculation loops of a BWR. The second nodalization models each of the ten jet pumps of each recirculation loop. The results obtained from the execution of both models are compared, using important variables such as pressures and mass costs for the same components of both models. In addition, the maximum pressure value generated on the pipe located upstream of the water hammer, relative to the design pressure of the pipe, is compared for each arrangement. (Author)

  9. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  10. TRAB, a transient analysis program for BWR. Part 1

    International Nuclear Information System (INIS)

    Rajamaeki, Markku.

    1980-03-01

    TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)

  11. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    programs that GE has participated in and describes the different options and approaches that have been used by various utilities in their design basis programs. Some of these variations deal with the scope and depth of coverage of the information, while others are related to the process (how the work is done). Both of these topics can have a significant effect on the program cost. Some insight into these effects is provided. The final section of the paper presents a set of lessons learned and a recommendation for an optimum approach to a design basis information program. The lessons learned reflect the knowledge that GE has gained by participating in design basis programs with nineteen domestic and international BWR owner/operators. The optimum approach described in this paper is GE's attempt to define a set of information and a work process for a utility/GE NSSS Design Basis Information program that will maximize the cost effectiveness of the program for the utility. (author)

  12. Comparative analysis of the simulation of the instantaneous closing of the discharge valve of a recirculation loop of a BWR with a model of recirculation loop with 2 jet pumps and another model with 20 jet pumps using RELAP5/SCDAPSIM Mod. 3.4; Analisis comparativo de la simulacion del cierre instantaneo de la valvula de descarga de un lazo de recirculacion de un BWR con un modelo de lazo de recirculacion con 2 bombas chorro y un modelo con 20 bombas chorro empleando RELAP5/SCDAPSIM Mod. 3.4

    Energy Technology Data Exchange (ETDEWEB)

    Araiza M, E.; Ortiz V, J.; Martinez C, E.; Amador G, R.; Castillo D, R., E-mail: enrique.araiza@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This work presents the results of the simulation of the instantaneous closing of the water hammer, of a recirculation loop using two different arrangements in the loops. One of these arrangements corresponds to the traditional model that uses only two jet pumps to simulate the twenty pumps of the two recirculation loops of a BWR. The second nodalization models each of the ten jet pumps of each recirculation loop. The results obtained from the execution of both models are compared, using important variables such as pressures and mass costs for the same components of both models. In addition, the maximum pressure value generated on the pipe located upstream of the water hammer, relative to the design pressure of the pipe, is compared for each arrangement. (Author)

  13. Cobra-TF simulation of BWR bundle dry out experiments

    Energy Technology Data Exchange (ETDEWEB)

    Frepoli, C.; Ireland, A.; Hochreiter, L.; Ivanov, K. [Penn State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, PA (United States); Velten, R. [Siemens Nuclear Power GmbH, Erlangen (Germany)

    2001-07-01

    The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. This study presents the results obtained with COBRA-TF for the simulation of the Siemens 9-9Q BWR Bundle Dryout experiments. The model includes 20 channels and 34 axial nodes in the heated section. The predicted critical power and dryout location is compared with the measured values. An assessment of the code entrainment and de-entrainment models is presented. (authors)

  14. A porous flow approach to model thermal non-equilibrium applicable to melt migration

    Science.gov (United States)

    Schmeling, Harro; Marquart, Gabriele; Grebe, Michael

    2018-01-01

    We develop an approach for heat exchange between a fluid and a solid phase of a porous medium where the temperatures of the fluid and matrix are not in thermal equilibrium. The formulation considers moving of the fluid within a resting or deforming porous matrix in an Eulerian coordinate system. The approach can be applied, for example, to partially molten systems or to brine transport in porous rocks. We start from an existing theory for heat exchange where the energy conservation equations for the fluid and the solid phases are separated and coupled by a heat exchange term. This term is extended to account for the full history of heat exchange. It depends on the microscopic geometry of the fluid phase. For the case of solid containing hot, fluid-filled channels, we derive an expression based on a time-dependent Fourier approach for periodic half-waves. On the macroscopic scale, the temporal evolution of the heat exchange leads to a convolution integral along the flow path of the solid, which simplifies considerably in case of a resting matrix. The evolution of the temperature in both phases with time is derived by inserting the heat exchange term into the energy equations. We explore the effects of thermal non-equilibrium between fluid and solid by considering simple cases with sudden temperature differences between fluid and solid as initial or boundary conditions, and by varying the fluid velocity with respect to the resting porous solid. Our results agree well with an analytical solution for non-moving fluid and solid. The temperature difference between solid and fluid depends on the Peclet number based on the Darcy velocity. For Peclet numbers larger than 1, the temperature difference after one diffusion time reaches 5 per cent of \\tilde{T} or more (\\tilde{T} is a scaling temperature, e.g. the initial temperature difference). Thus, our results imply that thermal non-equilibrium can play an important role for melt migration through partially molten systems

  15. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  16. Effect of Mars Atmospheric Loss on Snow Melt Potential in a 3.5 Gyr Mars Climate Evolution Model

    Science.gov (United States)

    Mansfield, Megan; Kite, Edwin S.; Mischna, Michael A.

    2018-04-01

    Post-Noachian Martian paleochannels indicate the existence of liquid water on the surface of Mars after about 3.5 Gya (Irwin et al., 2015, https://doi.org/10.1016/j.geomorph.2014.10.012; Palucis et al., 2016, https://doi.org/10.1002/2015JE004905). In order to explore the effects of variations in CO2 partial pressure and obliquity on the possibility of surface water, we created a zero-dimensional surface energy balance model. We combine this model with physically consistent orbital histories to track conditions over the last 3.5 Gyr of Martian history. We find that melting is allowed for atmospheric pressures corresponding to exponential loss rates of dP/dt∝t-3.73 or faster, but this rate is within 0.5σ of the rate calculated from initial measurements made by the Mars Atmosphere and Volatile EvolutioN (MAVEN) mission, if we assume all the escaping oxygen measured by MAVEN comes from atmospheric CO2 (Lillis et al., 2017, https://doi.org/10.1002/2016JA023525; Tu et al., 2015, https://doi.org/10.1051/0004-6361/201526146). Melting at this loss rate matches selected key geologic constraints on the formation of Hesperian river networks, assuming optimal melt conditions during the warmest part of each Mars year (Irwin et al., 2015, https://doi.org/10.1016/j.geomorph.2014.10.012; Kite, Gao, et al., 2017, https://doi.org/10.1038/ngeo3033; Kite, Sneed et al., 2017, https://doi.org/10.1002/2017GL072660; Stopar et al., 2006, https://doi.org/10.1016/j.gca.2006.07.039). The atmospheric pressure has a larger effect on the surface energy than changes in Mars's mean obliquity. These results show that initial measurements of atmosphere loss by MAVEN are consistent with atmospheric loss being the dominant process that switched Mars from a melt-permitting to a melt-absent climate (Jakosky et al., 2017, https://doi.org/10.1126/science.aai7721), but non-CO2 warming will be required if <2 Gya paleochannels are confirmed or if most of the escaping oxygen measured by MAVEN comes from H2O.

  17. Low-Li2O Frits: Selecting Glasses that Support the Melt Rate Studies and Challenge the Current Durability Model

    International Nuclear Information System (INIS)

    Peeler, D. K.; Edwards, T. B.

    2005-01-01

    During the progressive development of the cold cap model (as it applies to a potential melt rate predictive tool), the formation of an Al-Li-silicate phase was identified as an intermediate reaction phase that could possibly hinder melt rate for SB4. To test this theory, six glasses were designed (using Frit 320's composition as the baseline) to maintain a constant 20 wt% sum of alkali content (in frit) by varying Na 2 O to Li 2 O ratios. The Li 2 O concentration ranged from 8 wt% down to 0% in either 2% or 1% increments with the differences being accounted for by an increase in Na 2 O concentration. Although the primary objective of the ''lower Li 2 O'' frits was to evaluate the potential for melt rate improvements, assessments of durability (as measured by the Product Consistency Test (PCT)) were also performed. The results suggest that durable glasses can be produced with these ''lower Li 2 O'' frits should it be necessary to pursue this option for improving melt rate. In addition to the series of glasses to support melt rate assessments, a series of frits were also developed to challenge the current durability model based on the limits proposed by Edwards et al. (2004). Although the ''new'' limits allow access into compositional regions of interest (i.e., higher alkali systems) which can improve melt rate and/or waste loading, there may still be ''additional'' conservatism. In this report, two series of glasses were developed to challenge the ''new'' durability limits for the SB4 system. In the first series, the total alkali of the Frit 320-based glasses (designed to support the melt rate program) was increased from 20 wt% to 21 wt% (in the frit), but the series also evaluated the possible impact of various Na 2 O and Li 2 O mass ratio differences. The second series pushed the alkali limit in the frit even further with frits containing either 22 or 24 wt% total alkali as well as various Na 2 O and Li 2 O mass ratios. The results of the PCT evaluation indicated

  18. A study of heat capacity temperature limit of BWR

    International Nuclear Information System (INIS)

    Wang, Shih-Jen; Chen, Jyh-Jun; Chien, Chun-Sheng; Teng, Jyh-Tong

    2012-01-01

    Highlights: ► The purpose of this study is to verify the HCTL. ► MAAP4 was used as code to generate a realistic and convenient HCTL. ► The current HCTL curve causes confusing in reading data. ► The revised HCTL curves developed in this study. ► Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners’ group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

  19. A study of heat capacity temperature limit of BWR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shih-Jen, E-mail: sjenwang@iner.gov.tw [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Chen, Jyh-Jun [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China); Chien, Chun-Sheng [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer The purpose of this study is to verify the HCTL. Black-Right-Pointing-Pointer MAAP4 was used as code to generate a realistic and convenient HCTL. Black-Right-Pointing-Pointer The current HCTL curve causes confusing in reading data. Black-Right-Pointing-Pointer The revised HCTL curves developed in this study. Black-Right-Pointing-Pointer Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners' group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

  20. Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

    International Nuclear Information System (INIS)

    Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

    2004-01-01

    Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

  1. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

    1986-01-01

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  2. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Suzuki, H.; Hatamiya, S.; Murase, M.

    1988-01-01

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

  3. Modeling of the thermal physical process and study on the reliability of linear energy density for selective laser melting

    Directory of Open Access Journals (Sweden)

    Zhaowei Xiang

    2018-06-01

    Full Text Available A finite element model considering volume shrinkage with powder-to-dense process of powder layer in selective laser melting (SLM is established. Comparison between models that consider and do not consider volume shrinkage or powder-to-dense process is carried out. Further, parametric analysis of laser power and scan speed is conducted and the reliability of linear energy density as a design parameter is investigated. The results show that the established model is an effective method and has better accuracy allowing for the temperature distribution, and the length and depth of molten pool. The maximum temperature is more sensitive to laser power than scan speed. The maximum heating rate and cooling rate increase with increasing scan speed at constant laser power and increase with increasing laser power at constant scan speed as well. The simulation results and experimental result reveal that linear energy density is not always reliable using as a design parameter in the SLM. Keywords: Selective laser melting, Volume shrinkage, Powder-to-dense process, Numerical modeling, Thermal analysis, Linear energy density

  4. Dual Phase Lag Model of Melting Process in Domain of Metal Film Subjected to an External Heat Flux

    Directory of Open Access Journals (Sweden)

    Mochnacki B.

    2016-12-01

    Full Text Available Heating process in the domain of thin metal film subjected to a strong laser pulse are discussed. The mathematical model of the process considered is based on the dual-phase-lag equation (DPLE which results from the generalized form of the Fourier law. This approach is, first of all, used in the case of micro-scale heat transfer problems (the extremely short duration, extreme temperature gradients and very small geometrical dimensions of the domain considered. The external heating (a laser action is substituted by the introduction of internal heat source to the DPLE. To model the melting process in domain of pure metal (chromium the approach basing on the artificial mushy zone introduction is used and the main goal of investigation is the verification of influence of the artificial mushy zone ‘width’ on the results of melting modeling. At the stage of numerical modeling the author’s version of the Control Volume Method is used. In the final part of the paper the examples of computations and conclusions are presented.

  5. Eulerian fluid-structure analysis of BWR

    International Nuclear Information System (INIS)

    McMaster, W.H.

    1979-05-01

    A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

  6. Methyl Iodide Decomposition at BWR Conditions

    International Nuclear Information System (INIS)

    Pop, Mike; Bell, Merl

    2012-09-01

    Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

  7. Delivering high performance BWR fuel reliably

    Energy Technology Data Exchange (ETDEWEB)

    Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

    1998-07-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  8. PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

    OpenAIRE

    MELARA SAN ROMÁN, JOSÉ

    2016-01-01

    [EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

  9. Impact of metals recycling on a Swedish BWR decommissioning project

    International Nuclear Information System (INIS)

    Larsson, Arne; Lidar, Per; Hedin, Gunnar; Bergh, Niklas

    2014-01-01

    Decommissioning of nuclear power plants generates large volumes of radioactive or potentially contaminated metals. By proper management of the waste streams significant amounts can be free released and recycled either directly or after decontamination and melting. A significant part of the required work should be performed early in the process to make the project run smoothly without costly surprises and delays. A large portion of the clearance activities can be performed on-site. This on-site work should focus on the so called low-risk for contamination material. Other material can be decontaminated and released on site if schedule and the available facility areas so allow. It should be noted that the on-site decontamination and clearance activities can be a significant bottle neck for a decommissioning project. The availability of and access to a specialized metals recycling facility is an asset for a decommissioning project. This paper will describe the forecasted positive impact of a well-structured metals characterisation, categorisation and clearance process for a BWR plant decommissioning project. The paper is based on recent studies, performed projects and recent in-house development. (authors)

  10. Utility experience with BWR-PSMS

    International Nuclear Information System (INIS)

    Bond, G.R.

    1986-01-01

    The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

  11. Economic analysis of hydride fueled BWR

    International Nuclear Information System (INIS)

    Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

    2009-01-01

    The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

  12. Investigation of BWR stability in Forsmark 2

    International Nuclear Information System (INIS)

    Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

    1988-01-01

    A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  13. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  14. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1982-01-01

    A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  15. Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

    1996-01-01

    A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

  16. Numerical Modeling of Fluid Flow, Heat Transfer and Arc-Melt Interaction in Tungsten Inert Gas Welding

    Science.gov (United States)

    Li, Linmin; Li, Baokuan; Liu, Lichao; Motoyama, Yuichi

    2017-04-01

    The present work develops a multi-region dynamic coupling model for fluid flow, heat transfer and arc-melt interaction in tungsten inert gas (TIG) welding using the dynamic mesh technique. The arc-weld pool unified model is developed on basis of magnetohydrodynamic (MHD) equations and the interface is tracked using the dynamic mesh method. The numerical model for arc is firstly validated by comparing the calculated temperature profiles and essential results with the former experimental data. For weld pool convection solution, the drag, Marangoni, buoyancy and electromagnetic forces are separately validated, and then taken into account. Moreover, the model considering interface deformation is adopted in a stationary TIG welding process with SUS304 stainless steel and the effect of interface deformation is investigated. The depression of weld pool center and the lifting of pool periphery are both predicted. The results show that the weld pool shape calculated with considering the interface deformation is more accurate.

  17. Physics-Based Modeling of Electric Operation, Heat Transfer, and Scrap Melting in an AC Electric Arc Furnace

    Science.gov (United States)

    Opitz, Florian; Treffinger, Peter

    2016-04-01

    Electric arc furnaces (EAF) are complex industrial plants whose actual behavior depends upon numerous factors. Due to its energy intensive operation, the EAF process has always been subject to optimization efforts. For these reasons, several models have been proposed in literature to analyze and predict different modes of operation. Most of these models focused on the processes inside the vessel itself. The present paper introduces a dynamic, physics-based model of a complete EAF plant which consists of the four subsystems vessel, electric system, electrode regulation, and off-gas system. Furthermore the solid phase is not treated to be homogenous but a simple spatial discretization is employed. Hence it is possible to simulate the energy input by electric arcs and fossil fuel burners depending on the state of the melting progress. The model is implemented in object-oriented, equation-based language Modelica. The simulation results are compared to literature data.

  18. A model of sulphur solubility for hydrous mafic melts: application to the determination of magmatic fluid compositions of Italian volcanoes

    Directory of Open Access Journals (Sweden)

    M. Pichavant

    2005-06-01

    Full Text Available We present an empirical model of sulphur solubility that allows us to calculate f S2 if P, T, fO2 and the melt composition, including H2O and S, are known. The model is calibrated against three main experimental data bases consisting in both dry and hydrous silicate melts. Its prime goal is to calculate the f S2 of hydrous basalts that currently lack experimental constraints of their sulphur solubility behaviour. Application of the model to Stromboli, Vesuvius, Vulcano and Etna eruptive products shows that the primitive magmas found at these volcanoes record f S2 in the range 0.1-1 bar. In contrast, at all volcanoes the magmatic evolution is marked by dramatic variations in f S2 that spreads over up to 9 orders of magnitude. The f S2 can either increase during differentiation or decrease during decompression to shallow reservoirs, and seems to be related to closed versus open conduit conditions, respectively. The calculated f S2 shows that the Italian magmas are undersaturated in a FeS melt, except during closed conduit conditions, in which case differentiation may eventually reach conditions of sulphide melt saturation. The knowledge of f S2, fO2 and fH2O allows us to calculate the fluid phase composition coexisting with magmas at depth in the C-O-H-S system. Calculated fluids show a wide range in composition, with CO2 mole fractions of up to 0.97. Except at shallow levels, the fluid phase is generally dominated by CO2 and H2O species, the mole fractions of SO2 and H2S rarely exceeding 0.05 each. The comparison between calculated fluid compositions and volcanic gases shows that such an approach should provide constraints on both the depth and mode of degassing, as well as on the amount of free fluid in magma reservoirs. Under the assumption of a single step separation of the gas phase in a closed-system condition, the application to Stromboli and Etna suggests that the main reservoirs feeding the eruptions and persistent

  19. Petrological Geodynamics of Mantle Melting II. AlphaMELTS + Multiphase Flow: Dynamic Fractional Melting

    Science.gov (United States)

    Tirone, Massimiliano

    2018-03-01

    In this second installment of a series that aims to investigate the dynamic interaction between the composition and abundance of the solid mantle and its melt products, the classic interpretation of fractional melting is extended to account for the dynamic nature of the process. A multiphase numerical flow model is coupled with the program AlphaMELTS, which provides at the moment possibly the most accurate petrological description of melting based on thermodynamic principles. The conceptual idea of this study is based on a description of the melting process taking place along a 1-D vertical ideal column where chemical equilibrium is assumed to apply in two local sub-systems separately on some spatial and temporal scale. The solid mantle belongs to a local sub-system (ss1) that does not interact chemically with the melt reservoir which forms a second sub-system (ss2). The local melt products are transferred in the melt sub-system ss2 where the melt phase eventually can also crystallize into a different solid assemblage and will evolve dynamically. The main difference with the usual interpretation of fractional melting is that melt is not arbitrarily and instantaneously extracted from the mantle, but instead remains a dynamic component of the model, hence the process is named dynamic fractional melting (DFM). Some of the conditions that may affect the DFM model are investigated in this study, in particular the effect of temperature, mantle velocity at the boundary of the mantle column. A comparison is made with the dynamic equilibrium melting (DEM) model discussed in the first installment. The implications of assuming passive flow or active flow are also considered to some extent. Complete data files of most of the DFM simulations, four animations and two new DEM simulations (passive/active flow) are available following the instructions in the supplementary material.

  20. Melting temperature of graphite

    International Nuclear Information System (INIS)

    Korobenko, V.N.; Savvatimskiy, A.I.

    2001-01-01

    Full Text: Pulse of electrical current is used for fast heating (∼ 1 μs) of metal and graphite specimens placed in dielectric solid media. Specimen consists of two strips (90 μm in thick) placed together with small gap so they form a black body model. Quasy-monocrystal graphite specimens were used for uniform heating of graphite. Temperature measurements were fulfilled with fast pyrometer and with composite 2-strip black body model up to melting temperature. There were fulfilled experiments with zirconium and tungsten of the same black body construction. Additional temperature measurements of liquid zirconium and liquid tungsten are made. Specific heat capacity (c P ) of liquid zirconium and of liquid tungsten has a common feature in c P diminishing just after melting. It reveals c P diminishing after melting in both cases over the narrow temperature range up to usual values known from steady state measurements. Over the next wide temperature range heat capacity for W (up to 5000 K) and Zr (up to 4100 K) show different dependencies of heat capacity on temperature in liquid state. The experiments confirmed a high quality of 2-strip black body model used for graphite temperature measurements. Melting temperature plateau of tungsten (3690 K) was used for pyrometer calibration area for graphite temperature measurement. As a result, a preliminary value of graphite melting temperature of 4800 K was obtained. (author)

  1. Analysis of the thermal response of a BWR Mark-I containment shell to direct contact by molten core materials

    International Nuclear Information System (INIS)

    Kress, T.S.; Cleveland, J.C.

    1988-01-01

    This study was undertaken to evaluate the thermal response of a BWR Mark-I containment shell in the event of an accident severe enough for molten core materials to fall into the cavity beneath the rector vessel and eventually come into direct contact with the shell. An existing ORNL three-dimensional transient heat transport computer code, HEATING-6, was used for a specific 2-D case (and variations) for which representative melt/shell boundary conditions required as input were available from other studies. In addition to the use of HEATING-6, a simplified analytical steady-state correlation was developed and given the name BWR Liner Analysis Program (BWRLAP). BWRLAP was ''benchmarked'' by comparison with HEATING-6 and was then used to make a number of parametric calculations to investigate the sensitivities of the results to the inputs. 5 refs., 11 figs., 2 tabs

  2. Modeling of the thermal physical process and study on the reliability of linear energy density for selective laser melting

    Science.gov (United States)

    Xiang, Zhaowei; Yin, Ming; Dong, Guanhua; Mei, Xiaoqin; Yin, Guofu

    2018-06-01

    A finite element model considering volume shrinkage with powder-to-dense process of powder layer in selective laser melting (SLM) is established. Comparison between models that consider and do not consider volume shrinkage or powder-to-dense process is carried out. Further, parametric analysis of laser power and scan speed is conducted and the reliability of linear energy density as a design parameter is investigated. The results show that the established model is an effective method and has better accuracy allowing for the temperature distribution, and the length and depth of molten pool. The maximum temperature is more sensitive to laser power than scan speed. The maximum heating rate and cooling rate increase with increasing scan speed at constant laser power and increase with increasing laser power at constant scan speed as well. The simulation results and experimental result reveal that linear energy density is not always reliable using as a design parameter in the SLM.

  3. Modeling of spreading of the melted corium jet inside the pool of emergency heat removal during severe accidents at NPP

    Directory of Open Access Journals (Sweden)

    I. V. Kazachkov

    2012-03-01

    Full Text Available Important nuclear power safety problem in touch with modeling of melted corium jet spreading inside the coolant pool is considered in the paper. It appears by development of the passive protection systems against se-vere accidents. The non-linear mathematical developed model is presented for the jet under reactor vessel pool for one of the perspective passive protection systems and the results of its analysis and studies are given. The performed analysis and the results of the numerical simulation done on the base of the model have allowed estab-lishing the interesting behaviors of the system, which may be useful for the scientists, as well as the engineers-constructors of the passive protection systems against severe accidents.

  4. Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  5. Modelling of melting and solidification transport phenomena during hypothetical NPP severe accidents

    International Nuclear Information System (INIS)

    Sarler, B.

    1992-01-01

    A physical and mathematical framework to deal with the transport phenomena occuring during melting and solidification of the hypothetical NPP severe accidents is presented. It concentrates on the transient temperature, velocity, and species concentration distributions during such events. The framework is based on the Mixture Continuum Formulation of the components and phases, cast in the boundary-domain integral shape structured by the fundamental solution of the Laplace equation. The formulation could cope with various solid-liquid sub-systems through the inclusion of the specific closure relations. The deduced system of boundary-domain integral equations for conservation of mass, energy, momentum, and species could be solved by the boundary element discrete approximative method. (author) [sl

  6. Ex Situ Introduction and Distribution of Nonmetallic Particles in Aluminum Melt: Modeling and Experiment

    Science.gov (United States)

    Vorozhtsov, S.; Minkov, L.; Dammer, V.; Khrustalyov, A.; Zhukov, I.; Promakhov, V.; Vorozhtsov, A.; Khmeleva, M.

    2017-12-01

    Simulation results for mechanical mixing of aluminum melt using a mixing device are presented. It is found that an increase in the rotational velocity of the vortex device leads to intensive growth of turbulent diffusion in the liquid metal. Small vortices, although energetically weaker, can provide uniform mixing, creating regions of uniformity of several tens of microns. Experimental studies showed that use of a special mechanical mixing device enables injection of course titanium diboride particles and leads to significantly improved microstructure (reduction in average grain size from 900 μm to 400 μm) and mechanical properties of aluminum alloy A356. In particular, introduction of 0.2 wt.% titanium diboride particles using the vortex device leads to an increase in yield strength (from 15 MPa to 65 MPa) and ultimate tensile strength (from 140 MPa to 220 MPa) while preserving the ductility of the alloy ( 4.5%).

  7. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  8. GPE-BWR and the containment venting and filtering issue

    International Nuclear Information System (INIS)

    Palomo, J.; Santiago, J. de

    1988-01-01

    The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

  9. An analysis of instabilities of nuclear-coupled density-wave in BWR using modern frequency-domain control theory

    International Nuclear Information System (INIS)

    Zhao Yangping; Gao Huahun; Fu Longzhou

    1991-01-01

    A state-of-the-art multi-variable frequency-domain model has been developed for analysis of instabilities of nuclear-coupled density-wave in BWR core. The characteristic locus method is used for analysing the stability of BWR. A computer code-NUCTHIA has been derived. The model has been tested against the existing experimental data and compared with results of past single-variable analyses. By using the NUCTHIA code, the investigations of effects of main system parameters on BWW core stability have also been made. All the results are consistent with the experimental data

  10. Hot-melt extruded filaments based on pharmaceutical grade polymers for 3D printing by fused deposition modeling.

    Science.gov (United States)

    Melocchi, Alice; Parietti, Federico; Maroni, Alessandra; Foppoli, Anastasia; Gazzaniga, Andrea; Zema, Lucia

    2016-07-25

    Fused deposition modeling (FDM) is a 3D printing technique based on the deposition of successive layers of thermoplastic materials following their softening/melting. Such a technique holds huge potential for the manufacturing of pharmaceutical products and is currently under extensive investigation. Challenges in this field are mainly related to the paucity of adequate filaments composed of pharmaceutical grade materials, which are needed for feeding the FDM equipment. Accordingly, a number of polymers of common use in pharmaceutical formulation were evaluated as starting materials for fabrication via hot melt extrusion of filaments suitable for FDM processes. By using a twin-screw extruder, filaments based on insoluble (ethylcellulose, Eudragit(®) RL), promptly soluble (polyethylene oxide, Kollicoat(®) IR), enteric soluble (Eudragit(®) L, hydroxypropyl methylcellulose acetate succinate) and swellable/erodible (hydrophilic cellulose derivatives, polyvinyl alcohol, Soluplus(®)) polymers were successfully produced, and the possibility of employing them for printing 600μm thick disks was demonstrated. The behavior of disks as barriers when in contact with aqueous fluids was shown consistent with the functional application of the relevant polymeric components. The produced filaments were thus considered potentially suitable for printing capsules and coating layers for immediate or modified release, and, when loaded with active ingredients, any type of dosage forms. Copyright © 2016 Elsevier B.V. All rights reserved.

  11. Comparison of modelling and experimental results of anode surface melting by femtosecond laser-stimulated electrical discharges in small gaps

    International Nuclear Information System (INIS)

    Chen Jian; He Lingna; Farson, Dave F; Rokhlin, Stanislav I

    2011-01-01

    Experiments and particle-in-cell simulations of femtosecond laser-stimulated electrical discharges in submicrometre gaps between scanning tunnelling microscope tip cathodes and gold film anodes are described. In experiments at applied potentials of 35 V and less, discharges were detected either as self-terminating low-current pulses with durations less than 10 ns and magnitudes less than 200 mA or as higher-current, longer-duration current waveforms. The probability of occurrence of low-current pulses increased as applied potential was decreased, being certain at low potentials of 20-25 V. Low-current pulse waveforms and surface melting of gold anodes predicted by the simulations were compared with experiments. Laser stimulation was modelled by introducing partially ionized electrode materials into the simulation domain at a controlled rate. Simulation results showed that the duration of low-current pulses was influenced by the time over which material was added to the gap region, establishing the importance of electrode vaporization on discharge duration. Subsequently, partially ionized electrode materials were preloaded into the gap in controlled amounts in subsequent simulations. Peak currents predicted by these simulations were nearly equal to the low-current pulse measurements but simulated pulse durations were shorter than experiments. Thus, the time axis of simulation current profiles was normalized for equality of charge transfer with experiments. Anode temperatures and melt diameters calculated from normalized simulated heat input profiles were well matched to experimental measurements.

  12. Experimental studies on melt spreading, bubbling heat transfer, and coolant layer boiling

    International Nuclear Information System (INIS)

    Greene, G.A.; Finfrock, C.; Klages, J.; Schwarz, C.E.; Burson, S.B.

    1988-01-01

    Melt spreading studies have been undertaken to investigate the extent to which molten core debris may be expected to spread under gravity forces in a BWR drywell geometry. The objectives are to determine the extent of melt spreading as a function of melt mass,melt superheat, and water depth. These studies will enable an objective determination of whether or not core debris can spread up to and contact containment structures or boundaries upon vessel failure. Results indicate that the most important variables are the melt superheat and the water depth. Studies have revealed five distinct regimes of melt spreading ranging from hydrodynamically-limited to heat transfer-limited. A single parameter dimensionless correlation is presented which identified the spreading regime and allows for mechanistic calculation of the average thickness to which the melt will spread. 7 refs., 12 figs

  13. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    International Nuclear Information System (INIS)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

    2013-01-01

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  14. Prevention of organic iodide formation in BWR's

    International Nuclear Information System (INIS)

    Karjunen, T.; Laitinen, T.; Piippo, J.; Sirkiae, P.

    1996-01-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR's as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs

  15. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Monta, K.; Itoh, J.

    1992-01-01

    An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

  16. Rheological Model for Describing Viscometric Flows of Melts of Branched Polymers

    Czech Academy of Sciences Publication Activity Database

    Merzlikina, D.A.; Pyshnograi, G.; Pivokonský, Radek; Filip, Petr

    2016-01-01

    Roč. 89, č. 3 (2016), s. 652-659 ISSN 1062-0125 Grant - others:Russian Foundation for Basic Research(RU) 15-4-04003 Institutional support: RVO:67985874 Keywords : rheology * polymer melts * mesoscopic approach * rheological equation of state * viscometric fl ows Subject RIV: BK - Fluid Dynamics http://download.springer.com/static/pdf/8/art%253A10.1007%252Fs10891-016-1423-7.pdf?originUrl=http%3A%2F%2Flink.springer.com%2Farticle%2F10.1007%2Fs10891-016-1423-7&token2=exp=1481193683~acl=%2Fstatic%2Fpdf%2F8%2Fart%25253A10.1007%25252Fs10891-016-1423-7.pdf%3ForiginUrl%3Dhttp%253A%252F%252Flink.springer.com%252Farticle%252F10.1007%252Fs10891-016-1423-7*~hmac=cb34c324456883ecc6fc41d3feb0677836d7c454728829b5e085aac5ce7de417

  17. Modelling of melting and solidification transport phenomena during hypothetical NPP severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sarler, B [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    A physical and mathematical framework to deal with the transport phenomena occuring during melting and solidification of the hypothetical NPP severe accidents is presented. It concentrates on the transient temperature, velocity, and species concentration distributions during such events. The framework is based on the Mixture Continuum Formulation of the components and phases, cast in the boundary-domain integral shape structured by the fundamental solution of the Laplace equation. The formulation could cope with various solid-liquid sub-systems through the inclusion of the specific closure relations. The deduced system of boundary-domain integral equations for conservation of mass, energy, momentum, and species could be solved by the boundary element discrete approximative method. (author) [Slovenian] Predstavljeno je fizikalno in matematicno ogrodje za obravnavo prenosnih pojavov taljenja in strjevanja med hipoteticnimi tezkimi nezgodami v jedrskih elektrarnah. Osredotoceno je na popis neustaljene porazdelitve temperatur, hitrosti in koncentracij sestavin med taksnimi dogodki. Ogrodje temelji na formulaciji kontinuuma mesanice komponent in faz, v obliki robno obmocnih integralskih enacb, ki so sestavljena na podlagi fundamentalne resitve Laplace-ove enacbe. Formulacija lahko popisuje stevilne trdno-tekoce pod-sisteme na podlagi specificnih sklopitvenih relacij. Izpeljan sistem robno-obmocnih integralskih enacb za popis ohranitve mase, energije, gibalne kolicine in sestavin lahko resimo na podlagi diskretne aproksimativne metode robnih elementov. (author)

  18. Shallow fractionation signature of phase chemistry in Taburiente lavas, La Palma, Canary Islands: Results of MELTS modeling

    Science.gov (United States)

    Guetschow, H. A.; Nelson, B. K.

    2002-12-01

    Depth of crystal fractionation influences the chemical evolution of ocean island basalts and has significant implications for the physical structures of these volcanoes. In contrast to dominantly shallow systems such as Hawaii, a range of fractionation depths have been reported for Canary Islands lavas. Magmas erupted on La Palma preserve fluid- and melt-inclusion evidence for high-pressure (> 10 kbar) crystallization (Klügel et al., 1998; Hansteen et al., 1998; Nikogosian et al., 2002). If high-pressure fractional crystallization were an early and dominant process, it would generate specific patterns in rock and phase chemistry of eruptive sequences. Alkalic basalts from Taburiente volcano display coherent major element trends consistent with evolution dominated by fractional crystallization while their phenocryst compositions, trace elements, and isotopic trends require mixing between multiple sources. The current model confirms the importance of both fractionation and mixing to achieve the full range of lavas observed. A low-pressure (1 kbar) thermodynamic fractional crystallization model performed with the MELTS (Ghiorso and Sack, 1995) software closely reproduces major element trends from two stratigraphic sequences. This model also predicts the observed sequence of groundmass clinopyroxene compositions and phenocryst zoning reversals. In all low pressure simulations, olivine remains a modally significant liquidus phase during the first 20% and last 30% of the crystallization sequence, resulting in a negative correlation between the CaO and Fo content of olivine. These results are consistent with the presence of olivine phenocrysts that bear petrographic evidence of early crystallization, as well as observed compositional trends of groundmass olivine and clinopyroxene in Taburiente lavas. MELTS models that include an initial period of high pressure (12 kbar) clinopyroxene fractionation produce major element trends comparable to the low pressure model, but

  19. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  20. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  1. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    Alessandro Petruzzi; Shin Chin; Kostadin Ivanov; Asok Ray; Fan-Bill Cheung

    2005-01-01

    Full text of publication follows: The objective of the work presented in this paper research, which is supported by the US Department of Energy under the NEER program, is to develop an early anomaly detection methodology in order to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, based on the US NRC coupled code TRACE/PARCS, is being utilized as a generator of time series data for anomaly detection at an early stage. The concept of the methodology is based on the fact that nonlinear systems show bifurcation, which is a change in the qualitative behavior as the system parameters vary. Some of these parameters may change on their own accord and account for the anomaly, while certain parameters can be altered in a controlled fashion. The non-linear, non-autonomous BWR system model considered in this research exhibits phenomena at two time scales. Anomalies occur at the slow time scale while the observation of the dynamical behavior, based on which inferences are made, takes place at the fast time scale. It is assumed that: (i) the system behavior is stationary at the fast time scale; and (ii) any observable non-stationary behavior is associated with parametric changes evolving at the slow time scale. The goal is to make inferences about evolving anomalies based on the asymptotic behavior derived from the computer simulation. However, only sufficient changes in the slowly varying parameter may lead to detectable difference in the asymptotic behavior. The need to detect such small changes in parameters and hence early detection of an anomaly motivate the utilized stimulus-response approach. In this approach, the model

  2. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Peng, C.M.; Maly, J.

    1988-01-01

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  3. A coupled melt-freeze temperature index approach in a one-layer model to predict bulk volumetric liquid water content dynamics in snow

    Science.gov (United States)

    Avanzi, Francesco; Yamaguchi, Satoru; Hirashima, Hiroyuki; De Michele, Carlo

    2016-04-01

    Liquid water in snow rules runoff dynamics and wet snow avalanches release. Moreover, it affects snow viscosity and snow albedo. As a result, measuring and modeling liquid water dynamics in snow have important implications for many scientific applications. However, measurements are usually challenging, while modeling is difficult due to an overlap of mechanical, thermal and hydraulic processes. Here, we evaluate the use of a simple one-layer one-dimensional model to predict hourly time-series of bulk volumetric liquid water content in seasonal snow. The model considers both a simple temperature-index approach (melt only) and a coupled melt-freeze temperature-index approach that is able to reconstruct melt-freeze dynamics. Performance of this approach is evaluated at three sites in Japan. These sites (Nagaoka, Shinjo and Sapporo) present multi-year time-series of snow and meteorological data, vertical profiles of snow physical properties and snow melt lysimeters data. These data-sets are an interesting opportunity to test this application in different climatic conditions, as sites span a wide latitudinal range and are subjected to different snow conditions during the season. When melt-freeze dynamics are included in the model, results show that median absolute differences between observations and predictions of bulk volumetric liquid water content are consistently lower than 1 vol%. Moreover, the model is able to predict an observed dry condition of the snowpack in 80% of observed cases at a non-calibration site, where parameters from calibration sites are transferred. Overall, the analysis show that a coupled melt-freeze temperature-index approach may be a valid solution to predict average wetness conditions of a snow cover at local scale.

  4. Phenomenology of BWR fuel assembly degradation

    Science.gov (United States)

    Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

    2018-03-01

    Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

  5. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  6. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

    Maldonado, G. Ivan; Christenson, John M.; Renier, J.P.; Marcille, T.F.; Casal, J.

    2010-01-01

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

  7. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  8. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  9. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

    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)

  10. BWR-stability investigation at Forsmark 1

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

    1988-01-01

    A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  11. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  12. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  13. 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)

  14. LBB application in Swedish BWR design

    International Nuclear Information System (INIS)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-01-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions

  15. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  16. High-pressure anatectic paragneisses from the Namche Barwa, Eastern Himalayan Syntaxis: Textural evidence for partial melting, phase equilibria modeling and tectonic implications

    Science.gov (United States)

    Guilmette, C.; Indares, A.; Hébert, R.

    2011-05-01

    Rare kyanite-bearing anatectic paragneisses are found as boudins within sillimanite-bearing paragneisses of the core of the Namche Barwa Antiform, Tibet. In the present study, we document an occurrence from the NW side of the Yarlung Zangbo River. These rocks mainly consist of the assemblage garnet + K-feldspar + kyanite ± biotite + quartz + rutile ± plagioclase with kyanite locally pseudomorphed by sillimanite. The documented textures are consistent with the rocks having undergone biotite-dehydration melting in the kyanite stability field, under high-P granulite facies conditions, and having experienced melt extraction. However textures related to melt crystallization are ubiquitous both in polymineralic inclusions in garnet and in the matrix, suggesting that a melt fraction had remained in these rocks. Phase equilibria modelling was undertaken in the NCKFMASTHO system with THERMOCALC. P-T pseudosections built with the bulk compositions of one aluminous and one sub-aluminous paragneiss samples predict a biotite-kyanite-garnet-quartz-plagioclase-K-feldspar-liquid-rutile ± ilmenite field, in which biotite-dehydration melting occurs, located in the P-T range of ~ 800-875 °C and ~ 10-17 kbar. In addition, the topologies of these pseudosections are consistent with substantial melt loss during prograde metamorphism. A second set of P-T pseudosections with melt-reintegrated model bulk compositions were thus constructed to evaluate the effect of melt loss. The integration of textural information, precise mineral modes, mineral chemistry, and phase equilibria modelling allowed to constrain a P-T path where the rocks are buried to lower crustal depths at peak P-T conditions higher than 14 kbar and 825 °C, possibly in the order of 15-16 kbar and 850 °C, followed by decompression and cooling to P-T conditions of around 9 kbar and 810 °C, under which the remaining melt was solidified. The implications for granite production at the NBA and for Himalayan tectonic models

  17. Experimental Observation and Analytical Modeling of Melting and Solidification during Aluminum Alloy Repair by Turbulence Flow Casting

    Directory of Open Access Journals (Sweden)

    Muki Satya Permana

    2015-10-01

    Full Text Available This paper presents an overview on the state of the art of applicable casting technology for applications in the field of repairing aluminum alloy components. Repair process on the Al alloy sample using similar metal has been carried out to investigate the micro-structural effect. Joining occurs as a result of convection heat transfer of molten flow into the sand mold which melts the existing base metal inside the mold and subsequent solidification. The analytical model has been developed to describe aluminum alloy component repair by turbulence flow casting. The model is designed based on heat transfer principle that can handle the phenomena of heat flow. The experimental result and analytical model analyses pointed out that joint quality are greatly affected by parameters of preheating temperature and duration of molten metal flow in the mold. To obtain a desired metallurgical sound at the joint, the optimum temperature and time were adjusted in order to obtain a similarity of microstructure between filler and base metal. This model is aimed to predict the use of the process parameter ranges in order to have the optimum parameters when it is applied to the experiment. The fixed parameters are flow rate, sand ratio, and pouring temperature. The process parameters are preheating temperature and pouring time. It is concluded that analytical modeling has good agreement with the experimental result.

  18. Experimental Observation and Analytical Modeling of Melting and Solidification during Aluminum Alloy Repair by Turbulence Flow Casting

    Directory of Open Access Journals (Sweden)

    Muki Satya Permana

    2017-03-01

    Full Text Available This paper presents an overview on the state of the art of applicable casting technology for applications in the field of repairing Aluminium Alloy components. Repair process on the Aluminium sample using similar metal has been carried out to investigate the micro-structural effect. Joining occurs as a result of convection heat transfer of molten flow into the sand mold which melts the existing base metal inside the mold and subsequent solidification. The analytical model has been developed to describe aluminium component repair by Turbulence Flow Casting. The model built is based on heat transfer principle that can handle the phenomena of heat flow. The experimental result and analytical model analyses pointed out that joint quality are greatly affected by parameters of preheating temperature and duration of molten metal flow in the mold. To obtain a desired metallurgical sound at the joint, the optimum temperature and time were adjusted in order to obtain a similarity of microstructure between filler and base metal. This model is aimed to predict the use of the process parameter ranges in order to have the optimum parameters when it is applied to the experiment. The fixed parameters are flow rate, sand ratio, and pouring temperature. The process parameters are preheating temperature and pouring time. It is concluded that anaytical modeling has good agreement with the experimental result

  19. An integrated modeling system for estimating glacier and snow melt driven streamflow from remote sensing and earth system data products in the Himalayas

    Science.gov (United States)

    Brown, M. E.; Racoviteanu, A. E.; Tarboton, D. G.; Gupta, A. Sen; Nigro, J.; Policelli, F.; Habib, S.; Tokay, M.; Shrestha, M. S.; Bajracharya, S.; Hummel, P.; Gray, M.; Duda, P.; Zaitchik, B.; Mahat, V.; Artan, G.; Tokar, S.

    2014-11-01

    Quantification of the contribution of the hydrologic components (snow, ice and rain) to river discharge in the Hindu Kush Himalayan (HKH) region is important for decision-making in water sensitive sectors, and for water resources management and flood risk reduction. In this area, access to and monitoring of the glaciers and their melt outflow is challenging due to difficult access, thus modeling based on remote sensing offers the potential for providing information to improve water resources management and decision making. This paper describes an integrated modeling system developed using downscaled NASA satellite based and earth system data products coupled with in-situ hydrologic data to assess the contribution of snow and glaciers to the flows of the rivers in the HKH region. Snow and glacier melt was estimated using the Utah Energy Balance (UEB) model, further enhanced to accommodate glacier ice melt over clean and debris-covered tongues, then meltwater was input into the USGS Geospatial Stream Flow Model (GeoSFM). The two model components were integrated into Better Assessment Science Integrating point and Nonpoint Sources modeling framework (BASINS) as a user-friendly open source system and was made available to countries in high Asia. Here we present a case study from the Langtang Khola watershed in the monsoon-influenced Nepal Himalaya, used to validate our energy balance approach and to test the applicability of our modeling system. The snow and glacier melt model predicts that for the eight years used for model evaluation (October 2003-September 2010), the total surface water input over the basin was 9.43 m, originating as 62% from glacier melt, 30% from snowmelt and 8% from rainfall. Measured streamflow for those years were 5.02 m, reflecting a runoff coefficient of 0.53. GeoSFM simulated streamflow was 5.31 m indicating reasonable correspondence between measured and model confirming the capability of the integrated system to provide a quantification of

  20. An Integrated Modeling System for Estimating Glacier and Snow Melt Driven Streamflow from Remote Sensing and Earth System Data Products in the Himalayas

    Science.gov (United States)

    Brown, M. E.; Racoviteanu, A. E.; Tarboton, D. G.; Sen Gupta, A.; Nigro, J.; Policelli, F.; Habib, S.; Tokay, M.; Shrestha, M. S.; Bajracharya, S.

    2014-01-01

    Quantification of the contribution of the hydrologic components (snow, ice and rain) to river discharge in the Hindu Kush Himalayan (HKH) region is important for decision-making in water sensitive sectors, and for water resources management and flood risk reduction. In this area, access to and monitoring of the glaciers and their melt outflow is challenging due to difficult access, thus modeling based on remote sensing offers the potential for providing information to improve water resources management and decision making. This paper describes an integrated modeling system developed using downscaled NASA satellite based and earth system data products coupled with in-situ hydrologic data to assess the contribution of snow and glaciers to the flows of the rivers in the HKH region. Snow and glacier melt was estimated using the Utah Energy Balance (UEB) model, further enhanced to accommodate glacier ice melt over clean and debris-covered tongues, then meltwater was input into the USGS Geospatial Stream Flow Model (Geo- SFM). The two model components were integrated into Better Assessment Science Integrating point and Nonpoint Sources modeling framework (BASINS) as a user-friendly open source system and was made available to countries in high Asia. Here we present a case study from the Langtang Khola watershed in the monsoon-influenced Nepal Himalaya, used to validate our energy balance approach and to test the applicability of our modeling system. The snow and glacier melt model predicts that for the eight years used for model evaluation (October 2003-September 2010), the total surface water input over the basin was 9.43 m, originating as 62% from glacier melt, 30% from snowmelt and 8% from rainfall. Measured streamflow for those years were 5.02 m, reflecting a runoff coefficient of 0.53. GeoSFM simulated streamflow was 5.31 m indicating reasonable correspondence between measured and model confirming the capability of the integrated system to provide a quantification