WorldWideScience

Sample records for bwr boiling water

  1. The boiling water reactor BWR 90

    International Nuclear Information System (INIS)

    During the next decade a rise in the energy demand is expected worldwide, and this will in particular call for electricity generation capacity. A number of old generating plants, both nuclear and other plants, will probably have to be shut down for aging reasons, and their replacement will enhance the need for new generating capacity. The ABB Atom considers this situation to be met with a 'cautious evolution'. The offerings will largely be based on 'evolutions' of the successful light water reactor BWR 75. The new, evolutionary plant design of ABB Atom is the BWR 90. It can be designed, licensed and constructed in accordance with any safety regulations now in force or envisaged in the Western world. Emphasis has been, and will be, placed on features that facilitates licensing, shortens construction time and keeps electricity generation costs favourable. ABB also continues to develop a design of the 'passive' type, such as the 'passive' PIUS system, for possible deployment in the future. These efforts are more long-term activities, since development, verification and licensing of distinctly 'new' reactor concepts will have an extensive lead time. This paper presents the BWR 90 and its current status. The design is based on that of its forerunner, the BWR 75 standard design, taking into account the experiences gained from design and engineering, construction, commissioning, and operation of BWR 75 plants, the needs for adapting to new technologies and new safety requirements, as well as possibilities for simplifications and cost savings. (author) 4 figs

  2. Serious accidents on boiling water reactors (BWR)

    International Nuclear Information System (INIS)

    This short document describes, first, the specificities of boiling water reactors (BWRs) with respect to PWRs in front of the progress of a serious accident, and then, the strategies of accident management: restoration of core cooling, water injection, core flooding, management of hydrogen release, depressurization of the primary coolant circuit, containment spraying, controlled venting, external vessel cooling, erosion of the lower foundation raft by the corium). (J.S.)

  3. High Pressure Boiling Water Reactor HP-BWR

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  4. Radial nodalization effects on BWR [boiling water reactor] stability calculations

    International Nuclear Information System (INIS)

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

  5. BWR [boiling water reactor] shutdown margin model in SIMULATE-3

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) technical specifications require that the reactor be kept subcritical (by some prescribed margin) when at room temperature rodded conditions with any one control rod fully withdrawn. The design of an acceptable core loading pattern may require hundreds or thousands of neutronic calculations in order to predict the shutdown margin for each control rod. Direct, full-core, three-dimensional calculations with the SIMULATE-3 two-group advanced nodal code require 3 to 6 CPU min (on a SUN-4 workstation) for each statepoint/control rod that is computed. Such computing and manpower requirements may be burdensome, particularly during the early core design process. These requirements have been significantly reduced by the development of a fast, accurate shutdown margin model in SIMULATE-3. The SIMULATE-3 shutdown margin model achieves a high degree of accuracy and speed without using axial collapsing approximations inherent in many models. The mean difference between SIMULATE-3 one-group and two-group calculations is approximately - 12 pcm with a standard deviation of 35 pcm. The SIMULATE-3 shutdown margin model requires a factor of ∼15 less CPU time than is required for stacked independent two-group SIMULATE-3 calculations

  6. Plant operation performance improvements of the General Electric (GE) boiling water reactors (BWR'S)

    International Nuclear Information System (INIS)

    This paper summarizes some of the plant operation performance improvement techniques developed by the General Electric Company Nuclear Energy Business Operation for the General Electric Boiling Water Reactors (GE BWR's). Through the use of both thermal and plant hardware operating margins, substantial additional flexibility in plant operation can be achieved resulting in significant improvements in plant capacity and availability factor and potential fuel cycle economics for the currently operating or requisition GE BWR plants. This list of techniques includes expanding the BWR thermal power/moderator flow operating domain to the maximum achievable region, operation with a single recirculation loop out of service and operation at rated thermal power with reduced feedwater temperatures. These plant improvements and operating techniques can potentially increase plant capacity factor by 1% to 2% and provide additional fuel cycle economics savings to the GE BWR's owners

  7. Investigation of BWR [boiling water reactor] instability phenomena using RETRAN-03

    International Nuclear Information System (INIS)

    In 1988, LaSalle, a boiling water reactor (BWR)/5, experienced severe flux oscillations following a trip of both recirculation pumps. The flux oscillations were terminated by an automatic scram at 118% of rated neutron flux. As a result of this event, the U.S. Nuclear Regulatory Commission has asked the BWR utilities to develop procedural or hardware changes that will assure protection of all safety limits. The rapid growth of the oscillations at LaSalle, and the fact that previous stability analyses had predicted the plant to be very stable, emphasizes that a better understanding of this phenomenon is needed before the success of the long-term fixes can be assured. The intent of the Electric Power Research Institute's work was to use BWR transient methods to model reactor instabilities and investigate the factors that dominate this phenomenon. The one-dimensional transient code RETRAN-03 (Ref. 1) was used. The following conclusions are drawn: (1) RETRAN has demonstrated the ability to model BWR instability (nonlinear oscillations). (2) The general system behavior predicted by RETRAN in BWR stability analyses matches theoretical prediction and plant data. (3) These one-dimensional, time-domain results have increased the understanding of BWR stability phenomena and have helped optimize the long-term solutions being developed by the utilities

  8. Uncertainty calculation of emergency core cooling system for boiling water reactor (BWR-5)

    International Nuclear Information System (INIS)

    A brief description about uncertainty calculation of emergency core cooling system for boiling water reactor (BWR-5) is presented in this paper. Based on methodology of PSA level 1 and draft description of ECCS's document supplied by TOSHIBA (Code PSO-00-00097, July 2000) the event tree is built. The fault trees of three of subsystems HPCSS, LPCSS, LPCIS can be developed due to the simplified P and ID of ECCS and the reliability data accompanied. The computer code used to develop fault tree is KIRAP-tree and one used to find cut set and calculated uncertainty is KCUT. (author)

  9. A multi-cycle BWR [boiling water reactor] core reload design analysis system (MCAS)

    International Nuclear Information System (INIS)

    This paper describes the design, construction, and application of a software system (MCAS) for performing boiling water reactor reload core design analysis. MCAS provides for the execution of studies which analyze alternative reload strategies over a range of cycles. Studies are performed by preparing and executing sequential SIMULATE-E Haling depletions and storing the results on a data base for subsequent reporting and analysis. Application of MCAS has shown that the ability of efficiently and accurately predict the effects of next cycle design decisions on future cycles is a valuable capability. This capability results in the proper selection of BWR [boiling water reactor] reload fuel bundle enrichment and batch size as necessary for reload fuel supply planning and early identification and resolution of design problems which would prove expensive if discovered at a later time

  10. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    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

  11. Evaluation of a passive containment cooling system for a simplified BWR [boiling water reactor

    International Nuclear Information System (INIS)

    Simplified boiling water reactors (BWRs) are characterized for the adoption of a passive containment cooling system (PCCS) and a passive emergency core cooling system (ECCS). TOSPAC, which had been developed as the preliminary design code for several PCCS concepts, was compared with TRAC for verification. TOSPAC analyses were also performed to show the effectiveness of the isolation condenser (IC) as a PCCS over a wide range of break spectra. The selected reference plant for the analysis is a natural circulation BWR plant with 1,800-MW(thermal) power. The ECCS consists of a gravity-driven cooling system (GDCS) and depressurization valves. The IC and drywell cooler are considered for the PCCS. The IC units and drywell coolers are placed in the IC pool and GDCS pool, respectively

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

    Science.gov (United States)

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

    2014-09-01

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

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

    International Nuclear Information System (INIS)

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

  14. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

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

  15. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    The two-phase flow instability is of interest for the design and operation of many industrial systems such as boiling water reactors (BWRs), chemical reactors, and steam generators. In case of BWRs, the flow instabilities are coupled to the power instabilities via neutronic-thermal hydraulic feedbacks. Since these instabilities produce also local pressure oscillations, the coolant flashing plays a very important role at low pressure. Many frequency-domain codes have been used for two-phase flow stability analysis of thermal hydraulic industrial systems with particular emphasis to BWRs. Some were ignoring the effect of the local pressure, or the effect of 3D power oscillations, and many were not able to deal with the neutronics-thermal hydraulics problems considering the entire core and all its fuel assemblies. The new frequency domain tool uses the best available nuclear, thermal hydraulic, algebraic and control theory methods for simulating BWRs and analyzing their stability in either off-line or on-line fashion. The novel code takes all necessary information from plant files via an interface, solves and integrates, for all reactor fuel assemblies divided into a number of segments, the thermal-hydraulic non-homogenous non-equilibrium coupled linear differential equations, and solves the 3D, two-energy-group diffusion equations for the entire core (with spatial expansion of the neutron fluxes in Legendre polynomials).It is important to note that the neutronics equations written in terms of flux harmonics for a discretized system (nodal-modal equations) generate a set of large sparse matrices. The eigenvalue problem associated to the discretized core statics equations is solved by the implementation of the implicit restarted Arnoldi method (IRAM) with implicit shifted QR mechanism. The results of the steady state are then used for the calculation of the local transfer functions and system transfer matrices. The later are large-dense and complex matrices, (their size

  16. Water chemistry improvements in an operating boiling water reactor (BWR) and associated benefits

    International Nuclear Information System (INIS)

    Kernkraft Muhleberg (KKM) nuclear power plant is a BWR/4, the older of the two BWRs in Switzerland located in the outskirts of Bern. The plant is currently in its 37th year of continuous power operation, and has implemented major water chemistry improvements, including, hydrogen water chemistry (HWC), depleted zinc oxide (DZO) addition, NobleChem™, and On-Line NobleChem™ applications. In addition, the KKM plant has also performed other improvements such as maintaining low reactor water conductivity to mitigate intergranular stress corrosion crack (IGSCC) initiation and growth, as well as taking numerous actions to control radiation source term reduction. The actions taken to control the latter include replacement of the brass condenser tubes and an active cobalt source term reduction plan by eliminating the stellite control rod pins and rollers. These water chemistry improvements at the KKM plant have resulted in lower operating dose rates, lower drywell (shut down) dose rates and mitigation of shroud cracks. It is important to note that KKM is the only plant in the BWR industry that has monitored shroud internal diameter (ID) crack growth rates on a consistent basis using ultrasonic testing (UT) since 1993, thus providing an enormously valuable contribution to the BWR industry's in-plant crack growth rate data base. KKM plant has also installed tie rods in the shroud in 1996, an industry accepted approach. In addition, KKM also implemented NobleChem™ and On-Line NobleChem™ (OLNC) along with low hydrogen injection as additional proactive measures in 2000 and 2005 respectively to mitigate the growth of shroud cracks. There is reasonably clear evidence that since the implementation of OLNC, there is a consistent reduction in shroud crack growth rates showing mitigation of existing cracks. It is also evident that the drywell dose rates are showing a continuing decrease following 60Co source term reductions, DZO and OLNC implementations. This paper

  17. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O2; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-12-01

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

  19. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  20. Process inherent ultimate safety/boiling-water reactor PIUS/BWR

    International Nuclear Information System (INIS)

    This document is a series of viewgraphs on: design basis of PIUS/BWR, definition of PIUS/BWR, mechanisms of safe shutdown and afterheat cooling, advantages of PIUS/BWR, and research and development requirements

  1. Dynamic reconstruction and Lyapunov experiments from time series data in boiling water reactors. Application to B.W.R. stability analysis

    International Nuclear Information System (INIS)

    This paper shows how to obtain Lyapunov exponents from time series data on Boiling Water Reactor (BWR) stability. In order to validate the method, these characteristic exponents are compared with the ones obtained directly from the governing equations of the dynamic system. Finally, we present a method for obtaining the stability of the B.W.R. from Lyapunov exponents and describe some other applications related to limit cycles. (Author)

  2. Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  3. Standard- and extended-burnup PWR [pressurized-water reactor] and BWR [boiling-water reactor] reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs

  4. Application of reliability techniques to prioritize BWR [boiling water reactor] recirculation loop welds for in-service inspection

    International Nuclear Information System (INIS)

    In January 1988 the US Nuclear Regulatory Commission issued Generic Letter 88-01 together with NUREG-0313, Revision 2, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' to implement NRC long-range plans for addressing the problem of stress corrosion cracking in boiling water reactor piping. NUREG-0313 presents guidelines for categorizing BWR pipe welds according to their SCC condition (e.g., presence of known cracks, implementation of measures for mitigating SCC) as well as recommended inspection schedules (e.g., percentage of welds inspected, inspection frequency) for each weld category. NUREG-0313 does not, however, specify individual welds to be inspected. To address this issue, the Lawrence Livermore National Laboratory developed two recommended inspection samples for welds in a typical BWR recirculation loop. Using a probabilistic fracture mechanics model, LLNL prioritized loop welds on the basis of estimated leak probabilities. The results of this evaluation indicate that riser welds and bypass welds should be given priority attention over other welds. Larger-diameter welds as a group can be considered of secondary importance compared to riser and bypass welds. A ''blind'' comparison between the probability-based inspection samples and data from actual field inspections indicated that the probabilistic analysis generally captured the welds which the field inspections identified as warranting repair or replacement. Discrepancies between the field data and the analytic results can likely be attributed to simplifying assumptions made in the analysis. The overall agreement between analysis and field experience suggests that reliability techniques -- when combined with historical experience -- represent a sound technical basis on which to define meaningful weld inspection programs. 13 refs., 8 figs., 5 tabs

  5. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  6. Recriticality in a BWR [boiling water reactor] following a core damage event

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

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

  8. Transient boiling and void formation during postulated reactivity-initiated accident in BWR: Experimental simulation

    International Nuclear Information System (INIS)

    The current safety analysis of the postulated reactivity initiated accident (RIA) in the boiling water reactor (BWR) neglects the favorable effect of voids because of the difficulties in predicting void formation in transient boiling. This paper presents experimental results on the transient void formation in response to a step heating of a surface facing to low-pressure subcooled water. The void fractions are measured by measuring optically the water surface movement or water velocity induced by the void formation. (author)

  9. Transient boiling and void formation during postulated reactivity initiated accident in BWR: experimental simulation

    International Nuclear Information System (INIS)

    The current safety analysis of the postulated reactivity initiated accident (RIA) in the boiling water reactor (BWR) neglects the favorable effect of voids because of the difficulties in predicting void formation in transient boiling. This paper presents experimental results on the transient void formation in response to a step heating of a surface facing to low-pressure subcooled water. The void fractions are measured by measuring optically the water surface movement or water velocity induced by the void formation. (authors)

  10. High Pressure Boiling Water Reactor

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  13. Technical report on material selection and processing guidelines for BWR [boiling water reactor] coolant pressure boundary piping: Final report

    International Nuclear Information System (INIS)

    This report provides the technical bases for the NRC staff's revised recommended methods to control the intergranular stress corrosion cracking susceptibility of BWR piping. For piping that does not fully comply with the material selection, testing, and processing guideline combinations of this document, varying degrees of augmented inservice inspection are recommended. This revision also includes guidance and NRC staff recommendations (not requirements) regarding crack evaluation and weld overlay repair methods for long-term operation or for continuing interim operation of plants until a more permanent solution is implemented

  14. An assessment of BWR [boiling water reactor] Mark-II containment challenges, failure modes, and potential improvements in performance

    International Nuclear Information System (INIS)

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

  15. Recent developments in BWR water chemistry

    International Nuclear Information System (INIS)

    Water chemistry is of critical importance to the operation and economic viability of the Boiling Water Reactor (BWR). A successful water chemistry program will satisfy the following goals: - Minimize the incidence and growth of SCC/IASCC, - Minimize plant radiation fields controllable by chemistry, -Maintain fuel integrity by minimizing cladding corrosion, - Minimize flow-accelerated corrosion (FAC) in balance-of-plant components. The impact of water chemistry on each of these goals is discussed in more detail in this paper. It should be noted that water chemistry programs also include surveillance and operating limits for other plant water systems (e.g., service water, closed cooling water systems, etc.) but these are out of the scope of this paper. This paper reviews developments in water chemistry guidelines for U.S. BWR nuclear power plants. (author). 2 figs., 2 tabs., 7 refs

  16. Advances in the development and validation of CFD-BWR, a two-phase computational fluid dynamics model for the simulation of flow and heat transfer in boiling water reactors

    International Nuclear Information System (INIS)

    This paper presents recent advances in the validation of an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of two-phase flow and heat transfer phenomena in Boiling Water Reactor (BWR) fuel bundles. 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. We have described the model development strategy that has been adopted by the development team for the prediction of boiling flow regimes in a BWR fuel bundle. This strategy includes the use of local flow topology maps and flow topology specific phenomenological models. The paper reviews the key boiling phenomenological models and focuses on recent results of experiment analyses for the validation of two-phase BWR phenomena models including cladding-to-coolant heat transfer and Critical Heat Flux experiments and the BWR Full-size Assembly Boiling Test (BFBT). The two-phase flow models implemented in the CFD-BWR code can be grouped into three broad categories: models describing the vapor generation at the heated cladding surface, models describing the interactions between the vapor and the liquid coolant, and models describing the heat transfer between the fuel pin and the two-phase coolant. These models have been described and will be briefly reviewed. The boiling model used in the second generation of the CFD-BWR code includes a local flow topology map which allows the cell-by-cell selection of the local flow topology. Local flow topologies can range from a bubbly flow topology where the continuous phase is liquid, to a transition flow topology, to a droplet flow topology where the continuous phase is vapor, depending primarily on the local void fraction. The models describing the cladding-to-coolant heat transfer and the interplay between these models and the local flow topology are important in Critical Heat Flux (CHF) analyses, and will

  17. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    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

  18. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  19. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

  20. System comparative analysis of the most advanced pressured water reactors (PWR, WWER) and boiling water reactors (BWR) projects with the aim to choose the reactors for NPP construction in Kazakhstan

    International Nuclear Information System (INIS)

    organizations from 7 countries, SMART, integrated reactor, developed by Korea Atomic Energy Research Institute, Republic of Korea; CAREM, Argentina integrated reactor; MRX, integrated reactor, developed by Japan Atomic Energy Research Institute; UNITERM, NPP with integrated reactor, development by Research and development institute of power engineering (NIKIET), Russian Federation; AHEC-80, Russian NPP with integrated reactor, developed by OKB Mechanical Engineering (OKBM), Nizhny Novgorod, Russia. Moreover, following Boiling Water Reactor (BWR) projects have been subjected to the system comparative analysis. 1) Large Sized Reactors: ABWR, developed by Hitachi, Ltd, Japan, Toshiba Corporation, Japan ? G.E. USA; BWR-90, developed by Nuclear Systems Division, ABB Atom AB, Vasteras, Sweden; BWR-90+, developed by Nuclear Systems Division, ABB Atom AB, Vasteras, Sweden; SWR 1000, developed by Siemens Corporation, Germany; ESBWR, developed by General Electric Company, USA. 2) Medium Sized Reactors: SBWR, developed by General Electric Company, USA; HSBWR, developed by Hitachi Company, Ltd. 3) Small Sized Reactors: SSBWR, developed by Hitachi, Ltd, Japan; VK-300, BWR reactor, developed by Research and Development Institute of Power Engineering (NIKIET), Russia. Some data on the analysis of the condition and prospects of energy production and energy consumption, stations and networks in Kazakhstan are given. According to this analysis of nuclear power plants of average and low power are considered to be the most appropriate to construction in Kazakhstan. Recommendations on a choice of the most safe, reliable and economically competitive reactors have been made among the above-mentioned ones PWR, WWER and BWR for construction in Kazakhstan

  1. Droplet entrainment and deposition rate models for determination of boiling transition in BWR fuel assembly

    International Nuclear Information System (INIS)

    Droplet entrainment and deposition rates are of vital importance for mechanistic determination of critical power and location of boiling transition in a BWR fuel assembly. Data from high-pressure, high-temperature steam-water adiabatic experiments conducted in very tall test sections are used to develop a combination of equilibrium entrainment-deposition rate. Application of this combination to the heated tests conducted in a shorter test section of typical height of a BWR fuel assembly shows that correct split of total liquid in form of the film and droplets at the onset of annular-mist flow regime is also important to obtain good prediction of film flow rates/entrainment fraction. The improved model is then applied to simulate critical power tests in annulus and rod bundles. (author)

  2. Radiolysis of boiling water

    Science.gov (United States)

    Yang, Shuang; Katsumura, Yosuke; Yamashita, Shinichi; Matsuura, Chihiro; Hiroishi, Daisuke; Lertnaisat, Phantira; Taguchi, Mitsumasa

    2016-06-01

    γ-radiolysis of boiling water has been investigated. The G-value of H2 evolution was found to be very sensitive to the purity of water. In high-purity water, both H2 and O2 gases were formed in the stoichiometric ratio of 2:1; a negligible amount of H2O2 remained in the liquid phase. The G-values of H2 and O2 gas evolution depend on the dose rate: lower dose rates produce larger yields. To clarify the importance of the interface between liquid and gas phase for gas evolution, the gas evolution under Ar gas bubbling was measured. A large amount of H2 was detected, similar to the radiolysis of boiling water. The evolution of gas was enhanced in a 0.5 M NaCl aqueous solution. Deterministic chemical kinetics simulation elucidated the mechanism of radiolysis in boiling water.

  3. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    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

  4. Design certification program of the simplified boiling water reactor

    International Nuclear Information System (INIS)

    General Electric (GE), the US Department of Energy, the Electric Power Research Institute (EPRI), and utilities are undertaking a cooperative program to enable advanced light water reactor (ALWR) designs to be certified by the US Nuclear Regulatory Commission (NRC). GE is seeking to certify two advanced plants; the Advanced Boiling Water Reactor (ABWR) and the Simplified Boiling Water Reactor (SBWR). Both plants use advanced features that build on proven BWR technology

  5. Calculation of limit cycle amplitudes in commercial boiling water reactors

    International Nuclear Information System (INIS)

    This paper describes an investigation of the dynamic behavior of a boiling water reactor (BWR) in the nonlinear region corresponding to linearly unstable conditions. A nonlinear model of a typical BWR was developed. The equations underlying this model represent a one-dimensional void reactivity feedback, point kinetics with a single delayed neutron group, fuel behavior, and recirculation loop dynamics (described by a single-node integral momentum equation)

  6. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen

    International Nuclear Information System (INIS)

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  7. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    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

  8. Detonating gas in boiling water reactors

    International Nuclear Information System (INIS)

    The radiation in the core region of Boiling Water Reactors (BWRs) decomposes a small fraction of the coolant into hydrogen and oxygen, a phenomenon termed radiolysis. The radiolysis gas partitions to the steam during boiling. A 1000 MWe BWR produces around 1.5 tons of steam, containing 25 grams of radiolysis gas, per second. Practically all of the radiolysis gas is carried to the condenser and is taken care of by the condenser evacuation system and the off-gas system. The operation of these systems has been largely trouble-free. Radiolysis gas may also accumulate when stagnant steam condenses in pressurized pipes and components as a result of heat loss. Under certain circumstances a burnable mixture of hydrogen, oxygen and steam may form. Occasionally, the accumulated radiolysis gas has ignited. These incidents typically result in deformation of the components involved, but overpressure bursts have also occurred. Radiolysis gas accumulation in steam systems was largely overlooked by BWR designers (a likely technical reason for this is given in the report) and the problem had to be addressed by utilities. Even though the problem was recognized two decades ago, the counter-measures of today seem not always to be sufficient. Pipe-burst incidents in a German and a Japanese BWR recently attracted attention. Also, damage to a pilot valve in the steam relief system of a Swedish BWR forced a reactor shut-down during 2002. The recent incidents indicate that counter-measures against radiolysis gas accumulation in BWRs should be reviewed, perhaps also improved. The present report provides a short compilation of basic information related to radiolysis gas accumulation in BWRs. It is hoped that the compilation may prove useful to utilities and regulators reviewing the problem

  9. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

  10. Estimating boiling water reactor decommissioning costs. A user's manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    International Nuclear Information System (INIS)

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user's manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning

  11. Effects of lateral separation of oxidic and metallic core debris on the BWR [Boiling Water Reactor] MK I containment drywell floor

    International Nuclear Information System (INIS)

    In evaluating core debris/concrete interactions for a BWR MK I containment design, it has been common practice to assume that at reactor vessel breach, the core debris is homogeneous and of low viscosity so that it is uniformly distributed radially on the drywell floor. In a recent study performed by the NRC-sponsored Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, calculations indicate that at reactor vessel bottom head failure, the debris temperature is such that the metallic components (Zr, Fe, Ni, Cr) are completely molten while the oxidic components (UO2, ZrO2, FeO) are completely frozen. Thus, the frozen oxides are expected to remain within the reactor pedestal while the molten metallic species radially separate from the frozen oxidic species, flow through the opening in the reactor pedestal, and spread over the annular region of the drywell floor between the pedestal and the containment shell. This report assesses the impact on calculated debris gas release and the production and release of fission product-laden aerosols for two different cases of debris distribution: uniform distribution, and the laterally separated case of 95% oxides-5% metals inside the pedestal and 5% oxides-95% metals outside the pedestal. The computer codes used in this assessment are CORCON-MOD 2, MARCON 2.1B and VANESA

  12. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

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

  13. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

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

  14. Mitigation performance indicator for boiling water reactors

    International Nuclear Information System (INIS)

    All U.S. boiling water reactors (BWRs) inject hydrogen for mitigation of intergranular stress corrosion cracking (IGSCC), and most currently use or plan to use noble metals technology. The EPRI Boiling Water Reactor Vessels and Internals Project (BWRVIP) developed a Mitigation Performance Indicator (MPI) in 2006 to accurately depict to management the status of mitigation equipment and as a standardized way to show the overall health of reactor vessel internals from a chemistry perspective. It is a 'Needed' requirement in the EPRI BWR Water Chemistry Guidelines that plants have an MPI, and use of the BWRVIP MPI is a 'Good Practice'. The MPI is aligned with inspection relief criteria for reactor piping and internal components for U.S. BWRs. This paper discusses the history of the MPI, from its first use for plants operating with moderate hydrogen water chemistry (HWC-M) or Noble Metal Chemical Application (NMCA) + HWC to its more recent use for plants operating with On-Line NobleChem™ (OLNC) + HWC. Key mitigation parameters are discussed along with the technical bases for the indicators associated with the parameters. (author)

  15. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  16. TRAC-BD1: transient reactor analysis code for boiling-water systems

    Energy Technology Data Exchange (ETDEWEB)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented.

  17. SIMULATE-3K: Enhancements and Application to Boiling Water Reactor Transients

    International Nuclear Information System (INIS)

    The SIMULATE-3K (S-3K) reactor analysis code has been applied to a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) transients since 1993. Over the years, many changes have occurred in the S-3K channel hydraulics and ex-core component modeling. This paper summarizes those changes and outlines the status of existing vessel and steam line models. Examples are given for BWR transients that can be analyzed with S-3K

  18. Water chemistry management of nuclear power plant. Water chemistry management of BWR plant

    International Nuclear Information System (INIS)

    There are two kinds of nuclear power plants such as Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) in Japan. In this paper, a water chemistry management of BWR plant is explained. BWR plant makes steam produced in the reactor send to the turbine and produce power, then condensate in the main condenser and use again as feed water. The objects of water chemistry management of BWR are security of good conditions of fuel and structure materials and reduction of the dose equivalent and the radioactive waste. The volume of coolant depends on the temperature change, the concentration of boric acid for neutron absorber, lithium hydroxide for pH control and hydrogen gas for corrosion are controlled. Impurity metals in water of reactor are removed by the condensate demineralizer. The concentration of boron and lithium is controlled from 0 to 4000 ppm and from 0.2 to 2.2 ppm, respectively. On water chemistry technologies for dose reduction, oxygen injection into feed water and control operation of rate of Ni/Fe are explained. On the technologies for preventive maintenance, degassing operation of reactor and hydrogen injection into feed water are described. (S.Y.)

  19. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    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

  20. A parametric analysis of decay ratio calculations in a boiling water reactor model

    International Nuclear Information System (INIS)

    The results of an investigation of the effects of several parameters on the reactivity instability of a Boiling Water Reactor (BWR) calculational model are summarized. Calculations were performed for a typical BWR operated at low flow conditions, where reactivity instabilities are more likely to occur. The parameters investigated include the axial power shape (characterized by two separate parameters), the core pressure, and operating flow. All calculations were performed using the LAPUR code which was developed at the Oak Ridge National Laboratory for the dynamic modeling of large BWR's. 4 refs., 8 figs

  1. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  2. Boiling water reactor operator training and qualification in Japan

    International Nuclear Information System (INIS)

    Nuclear power plant operators in Japan are individuals employed by each electric power company. A recruit goes through his company's training; afterwards, he is given a qualification rating and is assigned to practical duty. The only formal qualification authorized by the Japanese government is the full-fledged shift supervisor. Other classifications such as assistant shift supervisor, shift foreman, reactor operator, and subreactor operator are all designated and appointed by each company's in-house regulations. As a part of the training system, power companies that require the use of a full-scope simulator in their training programs utilize the boiling water reactor (BWR) and pressurized water reactor operator training centers. Both were set up independently of the power companies. A synopsis of the BWR Operator Training Center Corp. (BTC ) and its training systems, features, performance evaluation, curriculum improvement, and related items is presented

  3. Overview of activities for the reduction of dose rates in Swiss boiling water reactors

    International Nuclear Information System (INIS)

    Since March 1990, zinc has been added to the reactor water of the boiling water reactor (BWR) Leibstadt (KKL) and, since January 1991, iron has been added to the BWR Muehleberg (KKM). These changes in reactor water chemistry were accompanied by a comprehensive R+D programme. This paper covers three selected topics: a) the statistical analysis of KKL reactor water data before and after zinc addition; b) the analysis of the KKL reactor water during the 1991 annual shutdown; c) laboratory autoclave tests to clarify the role of water additives on the cobalt deposition on austenitic steel surfaces. (author) 2 figs., 4 tabs

  4. Thermodynamic modeling of the processes in a boiling water reactor to buildup the magnetic corrosion product deposits

    International Nuclear Information System (INIS)

    Highlights: ► Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. ► Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe2O4]. ► GEM calculations applied to the boiling zone match with the EPMA and EXAFS findings. ► Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations. - Abstract: The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by using Gibbs Energy Minimization (GEM-Selector code) calculations of thermodynamic equilibrium at in situ temperatures and pressures. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe2O4] spinel solid solutions. GEM calculations applied to the boiling zone match with the electron probe microanalysis (EPMA) and Extended X-ray Absorption Fine Structure (EXAFS) findings, indicating that zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations under Zn water chemistry conditions. GEM results have helped to explain the existence of magnetic product deposits on the surface of the fuel element and the processes that take place in the reactor.

  5. Digital control application for the advanced boiling water reactor

    International Nuclear Information System (INIS)

    The Advanced Boiling Water Reactor (ABWR) is a 1300 MWe class Nuclear Power Plant whose design studies and demonstration tests are being performed by the three manufacturers, General Electric, Toshiba and Hitachi, under requirement specifications from the Tokyo Electric Power Company. The goals are to apply new technology to the BWR in order to achieve enhanced operational efficiencies, improved safety measures and cost reductions. In the plant instrumentation and control areas, traditional analog control equipment and wire cables will be replaced by distributed digital microprocessor based control units communicating with each other and the control room over fiber optic multiplexed data buses

  6. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

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

  7. Trends in BWR transient analysis

    International Nuclear Information System (INIS)

    While boiling water reactor (BWR) analysis methods for transient and loss of coolant accident analysis are well established, refinements and improvements continue to be made. This evolution of BWR analysis methods is driven by the new applications. This paper discusses some examples of these trends, specifically, time domain stability analysis and analysis of the simplified BWR (SBWR), General Electric's design approach involving a shift from active to passive safety systems and the elimination/simplification of systems for improved operation and maintenance

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

    International Nuclear Information System (INIS)

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

  9. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    Energy Technology Data Exchange (ETDEWEB)

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

  10. Industrial application of APOLLO2 to boiling water reactors

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

  12. Investigation of void effects in boiling water reactor fuels using neutron tomography

    OpenAIRE

    Loberg, John

    2006-01-01

    In a boiling water reactor (BWR), the void is correlated to dry out and the power level of the reactor. However, measuring the void is very difficult so it is therefore calculated with an accuracy that leaves room for improvements. Typically the uncertainty is ± 3% for 40% average void in the reactor. If the void could be determined with improved accuracy, both safety and economical features could be improved. X-ray tomography has previously been done on BWR fuel models in order to determine ...

  13. Uncommon water chemistry observations in modern day boiling water reactors

    International Nuclear Information System (INIS)

    Numerous technologies have been developed to mitigate intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) materials that include hydrogen water chemistry (HWC), noble metal chemical application (NMCA) and on-line NMCA (OLNC). These are matured technologies with extensive plant operating experiences, HWC – 32 years, NMCA – 18 years and OLNC – 9 years. Over the past three decades, numerous water chemistry data, dose rate data and IGSCC mitigation data relating to these technologies have been published and presented at many international conferences. However, there are many valuable and critical water chemistry and dose rate data that have gone unnoticed and unreported. The purpose of this paper is to highlight some of the uncommon water chemistry and dose rate experiences that reveal valuable information on the performance and durability of NMCA and OLNC technologies. Data will be presented, that have hitherto been unseen in public domain, from the lead OLNC plant in Switzerland giving reasons for some of the uncommon or overlooked water chemistry observations. They include, decreasing reactor water platinum concentration with each successive OLNC application, lack of increase in reactor water activation products in later applications, gradual disappearance of main steam line radiation (MSLR) monitor response decrease, Curium and Au-199 release during OLNC applications, rapid increase in reactor water clean-up conductivity, and Iodine, Mo-99 and Tc-99m spiking when hydrogen is interrupted and brought back to service, and main steam and reactor water conductivity spiking when clean-up beds or condensate demineralizers are changed. All these observations give valuable information on the success of OLNC applications and also signal the presence of sufficient noble metal on in-reactor surfaces from the long term durability and effectiveness stand point. Some of these observations can be used as secondary parameters, if and when a primary

  14. BWR internal cracking issues

    International Nuclear Information System (INIS)

    The regulatory issues associated with cracking of boiling water reactor (BWR) internals is being addressed by the Nuclear Regulatory Commission (NRC) staff and is the subject of a voluntary industry initiative. The lessons learned from this effort will be applied to pressurized water reactor (PWR) internals cracking issues

  15. 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).

  16. A semiempirical prediction of the decay ratio for the boiling water reactors start-up process

    International Nuclear Information System (INIS)

    During the start-up of a commercial boiling water reactor (BWR), the power and the coolant flow are continuously monitored. In order to prevent power instability events, the decay ratio (DR) could also be monitored. The process can be made safer if the operator could anticipate the DR too. DR depends on the power, the flow and many other quantities such as axial and radial neutron flux distribution, feed water temperature, void fraction, etc. A simple relationship for DR is derived. Three independent variables seem to be enough: the power, the flow and a single parameter standing for all other quantities which affect the DR. The relationship is validated with data from commercial BWR start-ups. A practical procedure for the start-up of a BWR is designed; it could help preventing instability events

  17. Description and status of BWR water radiolysis model

    International Nuclear Information System (INIS)

    A water radiolysis model for boiling water reactor primary coolant circuits has been developed jointly with Harwell Laboratories. Hydrogen water chemistry plant tests, on which the model is based, started in 1982. Since then, ten plant tests have been completed. Test results indicate that: (1) oxygen concentrations in the steam from plant to plant decreases fairly uniformly with increasing hydrogen concentration; (2) oxygen concentrations in the recirculating water varies widely with increasing hydrogen concentration in the feedwater; and (3) copper significantly suppresses the reduction of oxygen. The radiolysis model was developed to provide predictive conditions in the BWR primary circuit including the core. The objectives of the work are: (1) to develop and calibrate the model against experimental data; (2) to predict the concentration of various radiolytic species and oxygen suppression in various parts of the primary system under hydrogen water chemistry conditions; and (3) to define the role of copper on the reactions. The model considers the thermodynamics and kinetics of homogeneous chemical reactions in eight regions of the reactor and its external circuit. From initial concentrations, the model constructs a set of simultaneous equations to calculate hydrogen, hydrogen peroxide, and oxygen concentrations in various regions. Pilgrim data were used for the initial calibration. The model includes neutron and gamma dose rates in the core region and gamma dose rates in the other various regions, as well as temperature, power levels, heat transfer, steam quality, and gas/liquid transport in the core region, etc. The results of calculations for the oxygen concentrations in the recirculation water as a function of hydrogen concentration in the feedwater show good agreement with the experimental data obtained in a number of reactor tests

  18. Neutronic analysis and validation of boiling water reactor core designed by MCNPX code

    International Nuclear Information System (INIS)

    Highlights: • MCNPX code is used to design a model for BWR core. • The fuel enrichment is distributed in such a way to flat the power. • Validation of the BWR core model designed by MCNPX code. • Calculate Pu and its isotopes concentration at different burnup. - Abstract: This paper presents a design of boiling water reactor BWR model using MCNPX to develop benchmarks for checking the fuel management computer code packages. MCNPX code based on Monte Carlo method, is used to design a three dimensional model for BWR fuel assembly in typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. This design is used to study the thermal neutron flux and the pin by pin power distribution through the BWR core assemblies. The fuel used in BWR core is UO2 with three different types of enrichment (0.711%, 1.76% and 2.19%). This enrichment is distributed in such a way as to flatten the power. The effect of different enrichment values on the radial normalized power distribution is analyzed. The spent fuel in the reactor can be recycled, and plutonium and its isotopes can be extracted

  19. Dynamic simulation of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    For the application of modern control theory, specifically optimal control, to the boiling water reactor, it is necessary to have a linear model that is validated. The nonlinear model of the BWR derived on the basis of physical laws and empirical relations is linearized around an operating point and the model if verified against experimental results by simulating various tests such as the pressure transient test, change in power to recirculating pump etc. The transport delay occurring in the model is approximated by various representations and the results are compared with the exact delay representation. Validation such as discussed in the paper forms the basis for devising appropriate control strategies in the presence of disturbances. (author)

  20. Resolution of US regulatory issues involving boiling water reactor stability

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) and the Boiling Water Reactor Owners Group (BWROG) have been reexamining BWR instability characteristics and consequences since the March 1988 instability event at LaSalle Unit 2. The NRC and BWROG concluded that existing reactor protection systems do not prevent violation of the critical power ratio (CPR) safety limits caused by large asymmetric oscillations. The studies are also examining the need to modify the automatic and operator actions previously developed for response to an anticipated transient without scram (ATWS) event because of oscillation effects not fully considered in previous studies. This paper presents the current status of these studies and an assessment of actions needed to resolve the issue. (author)

  1. Multi-dimensional nodal analysis of boiling water reactor stability

    International Nuclear Information System (INIS)

    A computer program, NUFREQ-3D, was developed for boiling water reactor stability analysis. The code, which incorporates sophisticated thermal-hydraulic model coupled with a space dependent nodal neutronic model, is able to evaluate the system stabilities in terms of state variables such as inlet flow rate, power density, and system pressure. The detailed full 3-D representation was developed for more accurate stability analysis by using the sparse matrix techniques and by a channel grouping procedure. Results of modeling a representative operating BWR system show that spatial coupling has a significant effect on the prediction of stability margins. Comparisons of calculated transfer functions with the measured data also reveal that the code generally predict well the trends of system transfer functions

  2. Improvements in boiling water reactor designs and safety

    International Nuclear Information System (INIS)

    The advanced boiling water reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are discussed in this paper. They include: design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 last stage buckets; and advanced radwaste technology

  3. Boiling water reactor off-gas systems evaluation

    International Nuclear Information System (INIS)

    An evaluation of the off-gas systems for all 25 operating Boiling Water Reactors (BWR) was made to determine the adequacy of their design and operating procedures to reduce the probability of off-gas detonations. The results of the evaluations are that, of the 25 operable units, 13 meet all the acceptance criteria. The other 12 units do not have the features needed to meet the criteria, but have been judged to have, or are committed to provide, features which give reasonable assurance that the potential for external off-gas detonations is minimized. The 12 units which did not originally meet the criteria are aware of the potential hazards associated with off-gas detonations and have agreed to take action to minimize the probability of future detonations

  4. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    OpenAIRE

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use ...

  5. Evaluation of pressurized thermal shock in transitional condition for boiling water reactor pressure vessel

    International Nuclear Information System (INIS)

    The structural integrity for Pressurized Thermal Shock (PTS) was evaluated for the RPVs of Japanese Boiling Water Reactors (BWRs). It has been clarified that the BWR RPVs have the sufficient margin of fracture toughness by calculating the stress intensity factor in transitional condition and the acceptance criteria for RPV shell plate which is assumed to be neutron-irradiated in core region for 60 years. (author)

  6. Recycling heterogeneous americium targets in a boiling water reactor

    International Nuclear Information System (INIS)

    One of the limiting contributors to the heat load constraint for a long term spent fuel repository is the decay of americium-241. A possible option to reduce the heat load produced by Am-241 is to eliminate it via transmutation in a light water reactor thermal neutron environment, in particular, by taking advantage of the large thermal fission cross section of Am-242 and Am-242m. In this study we employ lattice loading optimization techniques to define the loadings and arrangements of fuel pins with blended americium and uranium oxide in boiling water reactor bundles, specifically, by defining the incineration of pre-loaded americium as an objective function to maximize americium transmutation. Subsequently, the viability of these optimized lattices is tested by assembling them into bundles with Am-spiked fuel pins and by loading these bundles into realistic three-dimensional BWR core-wide simulations that model multiple reload cycles and observe standard operational constraints. These simulations are possible via our collaboration with the Westinghouse Electric Co. which facilitates the use of industrial-caliber design tools such as the PHOENIX-4/POLCA-7 sequence and the Core Master 2 GUI work environment for fuel management. The resulting analysis confirms the ability to axially uniformly eliminating roughly 90% of the pre-loaded inventory of recycled Am-241 in BWR bundles with heterogeneous target pins. This high level of incineration was achieved within three to four 18-month operational cycles, which is equivalent to a typical in-core residence time of a BWR bundle.

  7. 21 CFR 872.6710 - Boiling water sterilizer.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling...

  8. Discussion on 'Electrochemical potential measurements under simulated BWR water chemistry conditions'

    International Nuclear Information System (INIS)

    In the above-referenced paper, Lin et al. report measurements of the corrosion potentials (the electrochemical potential or ECP) of types 304 and 316 SS in simulated boiling water reactor (BWR) heat transport environments at 270 C. There are four reasons for this discussion: to demonstrate that their theoretical explanation for the variation of ECP with oxygen concentration is inadequate; to show that their flow velocity/ECP results for oxygenated and hydrogenated systems are experimentally inconclusive because of experimental problems and, in any case, are inconsistent with electrochemical expectations; to cite previous work on the origin of the ECP of stainless steels in BWR environments that was not referenced in the paper but provides a basis for interpreting their data; and to identify previous work on the effect of Cu2+ on the ECP of type 304 SS, which was also not referenced in the paper

  9. Multivariable autoregressive model of the dynamics of a boiling water reactor

    International Nuclear Information System (INIS)

    An autoregressive (AR) model with pseudo-random binary sequence (PRBS) test signals was applied to the dynamics of the Japan Power Demonstration Reactor, a boiling water reactor (BWR). The decision of the order of the AR model was based on the Akaike criterion. Multi-input test signals of the PRBS were applied to the steam-flow control valve and the forced circulation pump speed control terminal. Seventeen variables including the instrumented fuel assemblies were observed. The AR model identification facilitated building the BWR dynamics model as a multivariable system. The experiment indicated that the BWR dynamics with rather intensive nonwhite noise interference was effectively represented by the AR model, which was compared with a linear theoretical dynamics model. The results suggested that the identified AR model plays an important role in verifying, modifying, and improving the theeoretical dynamics model

  10. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen; Estrategias de mitigacion de la corrosion intergranular en sistemas de reactores de agua en ebullicion (BWR). Accion combinada de la quimica del hidrogeno y del oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Verdugo, M.

    2015-07-01

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  11. Boiling of subcooled water in forced convection

    International Nuclear Information System (INIS)

    As a part of a research about water cooled high magnetic field coils, an experimental study of heat transfer and pressure drop is made with the following conditions: local boiling in tubes of small diameters (2 and 4 mm), high heat fluxes (about 1000 W/cm2), high coolant velocities (up to 25 meters/s), low outlet absolute pressures (below a few atmospheres). Wall temperatures are determined with a good accuracy, because very thin tubes are used and heat losses are prevented. Two regimes of boiling are observed: the establishment regime and the established boiling regime and the inception of each regime is correlated. Important delays on boiling inception are also observed. The pressure drop is measured; provided the axial temperature distribution of the fluid and the axial distributions of the wall temperatures, in other words the axial distribution of the heat transfer coefficients under boiling and non boiling conditions, at the same heat flux or the same wall temperatures, are taken in account, then total pressure drop can be correlated, but probably under certain limits of void fraction only. Using the same parameters, it seems possible to correlate the experimental values on critical heat flux obtained previously, which show very important effect of length and hydraulic diameter of the test sections. (authors)

  12. Water boiling kinetic in rapid decompression

    International Nuclear Information System (INIS)

    This study entering in the frame of a CEA, EDF and Framatome collaboration, has for objective to modelize two-phase flows in case of PWR Loca. The objective is to find, by taking in account the all imbalances, a formulation for the mass transfer at the interface water-vapor by the study of water boiling phenomenon in case of fast decompression such as a primary circuit break. In this accident, the estimation of boiling speeds in an essential parameter for determining the break discharge which conditions the safety systems design

  13. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    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

  14. Boiling Water Reactor Loading Pattern Optimization Using Simple Linear Perturbation and Modified Tabu Search Methods

    International Nuclear Information System (INIS)

    An automated system for designing a loading pattern (LP) for boiling water reactors (BWRs) given a reference LP and control rod (CR) sequence has been developed. This system employs the advanced nodal code SIMULATE-3 and a BWR LP optimization code FINELOAD-3, which uses a simple linear perturbation method and a modified Tabu search method to select potential optimized LP candidates. Both of these unique methods of FINELOAD-3 were developed to achieve an effective BWR LP optimization strategy and to have high computational efficiency. FINELOAD-3 also adjusts deep CR positions to compensate for the core reactivity deviation caused by fuel shuffling. The objective function is to maximize the end-of-cycle core reactivity while satisfying the specified thermal margins and cold shutdown margin constraints. This optimization system realized the practical application for real BWR LP design. Computer time needed to obtain an optimized LP for a typical BWR/5 octant core with 15 depletion steps is ∼4 h using an engineering workstation. This system was extensively tested for real BWR reload core designs and showed that the developed LPs using this system are equivalent or better than the manually optimized LPs

  15. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    International Nuclear Information System (INIS)

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

  16. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    The fuel assembly of a boiling water reactor contains a number of vertical fuel rods with their lower ends against a bottom tie plate. The rods are positioned by spacers, which are fixed to the canning. The upward motion is reduced by the top plate of a special design. (G.B.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

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

    International Nuclear Information System (INIS)

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

  19. The effectiveness of early hydrogen water chemistry on corrosion mitigation for boiling water reactors

    International Nuclear Information System (INIS)

    For mitigating intergranular stress corrosion cracking (IGSCC) in an operating boiling water reactor (BWR), the technology of hydrogen water chemistry (HWC) aiming at coolant chemistry improvement has been adopted worldwide. However, the hydrogen injection system is usually in an idle and standby mode during a startup operation. The coolant in a BWR during a cold shutdown normally contains a relatively high level of dissolved oxygen from intrusion of atmospheric air. Accordingly, the structural materials in the primary coolant circuit (PCC) of a BWR could be exposed to a strongly oxidizing environment for a short period of time during a subsequent startup operation. At some plants, the feasibility of hydrogen water chemistry during startup operations has been studied. It is technically difficult to directly procure water chemistry data at various locations of an operating reactor. Accordingly, the impact of startup operation on water chemistry in the PCC of a BWR operating under HWC can only be theoretically evaluated through computer modelling. In this study, a well-developed computer code DEMACE was used to investigate the variations in redox species concentration and in electrochemical corrosion potential (ECP) of components in the PCC of a domestic BWR during startup operations in the presence of HWC. Simulations were carried out for [H2]FWs ranging from 0.0 to 2.0 parts per million (ppm) and for power levels ranging from 3.8% to 11.3% during startup operations. Our analyses indicated that for power levels with steam generation in the core, a higher power level would tend to promote a more oxidizing coolant environment for the structural components and therefore lead to less HWC effectiveness on ECP reduction and corrosion mitigation. At comparatively lower power levels in the absence of steam, the effectiveness of HWC on ECP reduction was much better. The effectiveness of HWC in the PCC of a BWR during startup operations is expected to vary from location to

  20. Self-Sustaining Thorium Boiling Water Reactors

    OpenAIRE

    Ehud Greenspan; Jasmina Vujic; Francesco Ganda; Arias, Francisco J.

    2012-01-01

    A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR) proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorber...

  1. Boiling water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development, and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, Reactor Simulator Development: Workshop Material (2001). Course material for workshops using a WWER-1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21, 2nd edition, WWER-1000 Reactor Simulator: Workshop Material (2005). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator: Workshop Material (2005). This report consists of course material for workshops using a boiling water reactor (BWR) simulator

  2. CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

    International Nuclear Information System (INIS)

    At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

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

    Science.gov (United States)

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

    2016-09-01

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

  4. European simplified boiling water reactor (ESBWR) plant

    International Nuclear Information System (INIS)

    This paper covers innovative ideas which made possible the redesign of the US 660-MW Simplified Boiling Water Reactor (SBWR) Reactor Island for a 1,200-MW size reactor while actually reducing the building cost. This was achieved by breaking down the Reactor Island into multiple buildings separating seismic-1 from non-seismic-1 areas, providing for better space utilization, shorter construction schedule, easier maintainability and better postaccident accessibility

  5. U.S. experience with hydrogen water chemistry in boiling water reactors

    International Nuclear Information System (INIS)

    Hydrogen water chemistry in boiling water reactors is currently being adopted by many utilities in the U.S., with eleven units having completed preimplementation test programs, four units operating permanently with hydrogen water chemistry, and six other units in the process of installing permanent equipment. Intergranular stress corrosion cracking protection is required for the recirculation piping system and other regions of the BWR systems. The present paper explores progress in predicting and monitoring hydrogen water chemistry response in these areas. Testing has shown that impurities can play an important role in hydrogen water chemistry. Evaluation of their effects are also performed. Both computer modeling and in plant measurements show that each plant will respond uniquely to feedwater hydrogen addition. Thus, each plant has its own unique hydrogen requirement for recirculation system protecion. Furthermore, the modeling, and plant measurements show that different regions of the BWR respond differently to hydrogen injection. Thus, to insure protection of components other than the recirculation systems may require more (or less) hydrogen demand than indicated by the recirculation system measurements. In addition, impurities such as copper can play a significant role in establishing hydrogen demand. (Nogami, K.)

  6. Electrochemical potential measurements under simulated BWR water chemistry conditions

    International Nuclear Information System (INIS)

    Laboratory studies have been performed to investigate the stainless steel corrosion potential under simulated BWR coolant chemistry conditions. In addition to dissolved oxygen and hydrogen, test parameters also included chemical additives, metallic ions and hydrogen peroxide at various concentrations. The effect of water flow velocity was also investigated under various water chemistry conditions. The details of test results have been described elsewhere, and the highlights of the investigation are summarized in this paper. (J.P.N.)

  7. Boiling Water Reactor Fuel Assembly Axial Design Optimization Using Tabu Search

    International Nuclear Information System (INIS)

    In this paper the implementation of the tabu search (TS) optimization method to a boiling water reactor's (BWR's) fuel assembly (FA) axial design is described. The objective of this implementation is to test the TS method for the search of optimal FA axial designs. This implementation has been linked to the reactor core simulator CM-PRESTO in order to evaluate each design proposed in a reactor cycle operation. The evaluation of the proposed fuel designs takes into account the most important safety limits included in a BWR in-core analysis based on the Haling principle. Results obtained show that TS is a promising method for solving the axial design problem. However, it merits further study in order to find better adaptation of the TS method for the specific problem

  8. Evaluation of instrumentation for detection of inadequate core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    This report is a review of the Approach to Inadequate Core Cooling issue in Boiling Water Reactors (BWR). The report consists of seven sections. The principal conclusion is that the condition of the reference leg, and operator awareness of that condition are of primary importance in level indication reliability for safety. An indication of reference leg level and temperature displayed to the operators would be a useful enhancement of reliability and a guide to further operator action in all circumstances. We conclude that the BWR practice of multiple, redundant coolant level measurements, with overlapping ranges, can be a reliable basis for indication of approach to an ICC condition, and, in correlation with the other control and safety systems of modern BWRs, will prevent unsafe conditions

  9. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    A boiling water reactor fuel assembly is described which has vertical fuel rods and guide tubes positioned below the fuel rods and receiving control rod fingers and acting as water pipes, the guide tubes each being formed of a plurality of parts including a part secured to a grid plate positioned in the fuel assembly container, and low parts which fit into holes formed in the bottom of the fuel assembly. There is a flexible connection between the upper and lower parts of the guide tubes to allow for a certain tolerance in the procedure of manufacturing the various parts to allow insertion of the fuel rod bundle into the fuel assembly container

  10. Scaling laws and design aspects of a natural-circulation-cooled simulated boiling water reactor fuel assembly

    International Nuclear Information System (INIS)

    In order to study the thermohydraulic behavior of a natural-circulation-cooled boiling water reactor (BWR) fuel assembly, such as void drift, flow pattern distribution, and stability, a scaled loop geometry is designed. For modeling the steam/water flow in a BWR fuel assembly, scaling criteria are derived using the one-dimensional drift-flux model. Thermal equilibrium and subcooled boiling conditions are treated separately, resulting in one overall set of criteria. Scaling on all flow regimes that can be present in a normal fuel assembly leads to fixing both the assembly mass flux and the geometric dimensions. When Freon-12 is used as a modeling fluid, model assembly dimensions must be 0.46 of the prototype. Total power consumption must be reduced by a factor 50. To sustain cooling by natural circulation, a modeled chimney and downcomer are included

  11. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de inyeccion de agua de refrigeracion a baja presion (LPCI) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C., E-mail: renedelgado2015@hotmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

  12. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

  13. TARMS2, New-generation BWR core management system

    International Nuclear Information System (INIS)

    Toshiba has developed a prototype of a new-generation core management system for the boiling water reactor (BWR) TARMS2. It contains function modules for core monitoring and core prediction analysis. TARMS2 is equipped with an advanced three-dimensional BWR core physics model, LOGOS

  14. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

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

  15. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

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

  16. Neutronic challenges of advanced boiling water reactor designs

    International Nuclear Information System (INIS)

    The advancement of Boiling Water Reactor technology has been under investigation at the Center for Advance Nuclear Energy Systems at MIT. The advanced concepts under study provide economic incentives through enabling further power uprates (i.e. increasing vessel power density) or better fuel cycle uranium utilization. The challenges in modeling of three advanced concepts with focus on neutronics are presented. First, the Helical Cruciform Fuel rod has been used in some Russian reactors, and studied at MIT for uprating the power in LWRs through increased heat transfer area per unit core volume. The HCF design requires high fidelity 3D tools to assess its reactor physics behavior as well as thermal and fuel performance. Second, an advanced core design, the BWR-HD, was found to promise 65% higher power density over existing BWRs, while using current licensing tools and existing technology. Its larger assembly size requires stronger coupling between neutronics and thermal hydraulics compared to the current practice. Third is the reduced moderation BWRs, which had been proposed in Japan to enable breeding and burning of fuel as an alternative to sodium fast reactors. Such technology suffers from stronger sensitivity of its neutronics to the void fraction than the traditional BWRs, thus requiring exact modeling of the core conditions such as bypass voiding, to correctly characterize its performance. (author)

  17. Present status of maintenance technologies for boiling-water-reactor power plants

    International Nuclear Information System (INIS)

    Toshiba places the highest priority on maintenance technologies for boiling-water-reactor (BWR) power plants. These activities are based on our motto, 'Ensuring stable operation of BWRs throughout the plant life cycle'. A quarter of a century has passed since the construction of the first such plant in which Toshiba was involved, and preventive maintenance is therefore a matter of great importance for BWRs. This paper presents an overview of plant monitoring and diagnosis, preventive maintenance of equipment, and ensuring the high quality of plant improvement or annual inspection work. (author)

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

    OpenAIRE

    Alejandro Nuñez-Carrera; Raúl Camargo-Camargo; Gilberto Espinosa-Paredes; Adrián López-García

    2012-01-01

    The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the...

  19. The Nuclear option for U.S. electrical generating capacity additions utilizing boiling water reactor technology

    International Nuclear Information System (INIS)

    The technology status of the Advanced Boiling Water (ABWR) and Simplified Boiling Water (SBWR) reactors are presented along with an analysis of the economic potential of advanced nuclear power generation systems based on BWR technology to meet the projected domestic electrical generating capacity need through 2005. The forecasted capacity needs are determined for each domestic North American Electric Reliability Council (NERC) region. Extensive data sets detailing each NERC region's specific generation and load characteristics, and capital and fuel cost parameters are utilized in the economic analysis of the optimal generation additions to meet this need by use of an expansion planning model. In addition to a reference case, several sensitivity cases are performed with regard to capital costs and fuel price escalation

  20. Experiences with monitoring and control of microbiological growth in the standby service water system of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    In 1989, the Unites States Nuclear Regulatory Commission formally recognized the potential for nuclear accidents resulting from microbiological causes. Such causes range from loss of heat transfer due to microbiological fouling, to loss of system integrity caused by microbiologically influenced corrosion (MIC). As a result of these potential problems, monitoring, mitigation, and control procedures must be developed by all regulated plants. In developing a control and mitigation strategy for the standby service water system of a boiling water reactor (BWR) nuclear power plant, numerous monitoring techniques were employed to evaluate effectiveness. This paper describes the monitoring techniques that were evaluated, and those that ultimately proved to be effective

  1. Fuel recycling in boiling water reactors

    International Nuclear Information System (INIS)

    The present study confirms the feasibility of inserting mixed-oxid-fuel assemblies (MOX-FA) in boiling-water reactors in conjunction with reactivity-equivalent uranium-fuel assemblies. First, the established calculation methods were extended according to the specific MOX-uranium mutual interaction effects. Then, typical bundle-structures were analysed according to their neutron-physical features. The reactor-simulations show a non-critical behaviour with respect to limiting conditions and reactivity control. The variation of the isotopic composition and the plutonium content with its effects on the physical features was considered. (orig.) With 6 refs., 3 tabs., 29 figs

  2. Simulation of Boiling Water Reactor dynamics

    International Nuclear Information System (INIS)

    This master thesis describes a mathematical model of a boiling water reactor and address the dynamic behaviour of the neutron kinetics, boilding dynamics and pressur stability. The simulation have been done using the SIMNON-program. The meaning were that the result from this work possibly would be adjust to supervision methods suitable for application in computer systems. This master thesis in automatic control has been done at the Department of Automatic Control, Lund Institute of Technology. The initiative to the work came from Sydkraft AB. (author)

  3. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    A fuel assembly for a boiling water reactor comprises a plurality of fuel rods which constitute four partial bundles and are surrounded by a fuel channel system comprising one partial tube for each partial bundle. Each of the four partial bundles rests on a bottom tie plate and is positioned with respect to the others by means of a common top tie plate which is provided with a lifting loop which is sufficiently strong to be able to lift the four partial bundles simultaneously, a major part of the lifting force being transmitted to said bottom tie plates via a plurality of supporting fuel rods

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

    International Nuclear Information System (INIS)

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

  5. Analytical modeling of instabilities of two-phase flows in BWR boiling channels

    International Nuclear Information System (INIS)

    New approaches are presented in studying the mechanism of the coupled thermal/nuclear instabilities in BWR's. For the case of instabilities of the in-phase mode, a simple model based on the point-reactor model for the neutron dynamics and the lumped parameter model for the fuel dynamics was found useful in revealing the basic mechanism. For the case of instabilities of the out-of-phase mode, new models are developed which, in different levels of sophistication, take into consideration the nuclear/thermal interaction between different parts of a BWR core. In a simpler model called the subcore model, a coefficient is introduced which allows the neutron density in one subcore to be affected by the reactivity in another subcore. In a more completed model called the multi-point reactor model, a similar coefficient is introduced in establishing the neutron dynamics equation for each subcore, which directly accounts for the effect of neutron diffusion between subcores. Results are presented for these models, and discussions are made on their usefulness. (author)

  6. Modeling of two-phase flow in boiling water reactor using phase-weighted ensemble average method

    International Nuclear Information System (INIS)

    Investigations into boiling, the generation of vapor and the prediction of its behavior are important in the stability of boiling water reactors. The present models are limited to simplifications made to draw governing equations or lack of closure framework of the constitutive relations. The commercial codes fall into this category as well. Consequently, researchers cannot simply find the comprehensive updated relations before simplification in order to simplify them for their own works. This study offers a state of the art, phase-weighted, ensemble-averaged, two-phase flow, two-fluid model for the simulation of two-phase flow with heat and mass transfer. This approach is then used for modeling the bulk boiling (thermal-hydraulic modeling) in boiling water reactors. The resultant approach is based on using the energy balance equation to find a relation for quality of vapor at any point. The equations are solved using SIMPLE algorithm in the finite volume method and the results compared with real BWR (PB2 BWR/4 NPP) and the boiling data. Comparison shows that the present model is satisfactorily improved in accuracy.

  7. Boils

    Science.gov (United States)

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

  8. Nuclear power plant with pressure vessel boiling water reactor VK-300 for district heating and electricity supply

    International Nuclear Information System (INIS)

    The viability for Russia of the Boiling Water Reactor (BWR) concept has been shown by a number of feasibility studies fulfilled for perspective sites with increased energy demands. Russia has long (31 year) successful experience in operation of NPPs with the vessel-type boiling reactor VK-50 which is located in the city of Dimitrovgrad. Taking into account the large Russian district heating market, it is expedient to apply this concept (BWR) not only for electricity supply, but also for district heating. This is a way to increase of nuclear power plant competitiveness along with good safety performance. The safety and protection of nuclear heat customer is guaranteed by reliable technical means which are well checked at Russian nuclear sites. (author)

  9. Boiled Water Temperature Measurement System Using PIC Microcontroller

    OpenAIRE

    A.T.KARUPPIAH, AZHA. PERIASAMY, P.RAJKUMAR

    2013-01-01

    The measurement system for temperature of boiled water is a critical task in industry. In this paper we designed and implemented a PIC micro controller based boiled water temperature measurement system using PIC 18F452 and national semiconductors LM35 temperature sensor. The designing system is used to measure the tank I boiled water temperature value. If the temperature value reaches the set value high temperature relay board becomes ON to control the solenoid valve. The high temperature of ...

  10. Boiling in the presence of boron compounds in light water reactors

    International Nuclear Information System (INIS)

    The scope of the thesis on boiling in the presence of boron compounds in light water reactors was to study the effects of the boron compound addition on the heat removal from the fuel elements. For an effective cooling of the fuel elements in case of boiling processes a high heat transfer coefficient is of importance. Up to now experimental studies were not performed under reactor specific conditions, for instance with respect to the geometry of the flow conditions, high temperature and pressure levels were not represented. Therefore the experiments in the frame of the thesis were using reactor specific parameters. The test facility SECA (study into the effects of coolant additives) was designed and constructed. The experiments simulated the conditions of normal PWR operation, accidental PWR and accidental BWR conditions.

  11. Corrosion products release from steel surface into BWR water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kritsky, V.G.; Korolev, A.S.; Berezina, I.G.; Sofyin, M.V.

    1986-02-01

    Factors influencing steel corrosion product release and transfer into a BWR primary circuit have been studied and reported on in this paper. The study of corrosion kinetics and corrosion product release was carried out on the samples tested under RBMK NPP condensate-feedwater cycle conditions, as well as, under test rig conditions. The ratio of corrosion product specific mass, transferred to the water, to the whole corrosion product specific mass of steel, formed under the given conditions was determined and used as a criterion, characterizing the extent of corrosion product transfer from the steel surface into the water.

  12. Water Boiling inside Carbon Nanotubes: Towards Efficient Drug Release

    OpenAIRE

    Chaban, Vitaly V.; Prezhdo, Oleg V.

    2012-01-01

    We show using molecular dynamics simulation that spatial confinement of water inside carbon nanotubes (CNT) substantially increases its boiling temperature and that a small temperature growth above the boiling point dramatically raises the inside pressure. Capillary theory successfully predicts the boiling point elevation down to 2 nm, below which large deviations between the theory and atomistic simulation take place. Water behaves qualitatively different inside narrow CNTs, exhibiting trans...

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

    International Nuclear Information System (INIS)

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

  14. 44 BWR Waste Package Loading Curve Evaluation

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU

  15. 44 BWR Waste Package Loading Curve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2001-11-05

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU.

  16. Stability monitoring of the Dodewaard boiling-water reactor

    International Nuclear Information System (INIS)

    Methods for measuring the stability of a boiling-water are discussed. The results of experiments performed on the Dodewaard reactor (The Netherlands) are reported. Research on this reactor is of interest as it is cooled by natural circulation, a cooling principle that is also being considered for new reactor design. The stability of the Dodewaard reactor was studied both with deterministic methods (control-rod steps and pressure-valve movements) and by noise analysis. The latter method can be applied during normal operation and avoids any intentional system disturbance. Reactorkinetic stability, thermal-hydraulic stability and total-plant stability were investigated separately. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced; it was tested thorougly. It can be derived on-line from the noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations was calculated in order to assure a proper stability surveillance. A novel technique is presented, which enables the variations of the in-core coolant velocity to be determined by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies were performed on the fuel time constant, a parameter of great importance to the reactor-kinetic stability. It is shown that the effective value of this constant can be much smaller than the value commonly agreed on (author). 71 figs.; 73 figs,; 21 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are 2-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown. (orig.)

  19. Technique for technological calculation of critical flow of boiling water

    International Nuclear Information System (INIS)

    Average values of friction factor and mach number for a critical flow of boiling water are determined on the basis of computerized processing of experimental data. Empirical formula, relating these values, which can be used for technological calculations of critical conditions of boiling water flow through transport pipelines, is derived

  20. Final air test results for the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    A loss-of-coolant accident (LOCA) in a boiling-water reactor (BWR) power plant has never occurred. However, because this type of accident is particularly severe, it is used as a principal basis for design. During a hypothetical LOCA in a Mark I BWR, air followed by steam is injected from a drywell into a toroidal wetwell about half-filled with water. A series of consistent, versatile, and accurate air-water tests simulating LOCA conditions was completed in the Lawrence Livermore Laboratory 1/5-Scale Mark I BWR Pressure Suppression Experimental Facility. Results from this test series were used to quantify the vertical loading function and to study the associated fluid dynamic phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variations of hydrodynamic-generated vertical loads with changes in drywell pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1/5-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings which are invariant. These groupongs show that if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor; the time when these forces occur is reduced by the square root of the scale factor

  1. Stability monitor for Boiling Water Reactors based on the Multivariate Empirical Mode Decomposition

    International Nuclear Information System (INIS)

    Highlights: • Stability monitor based on a novel technique was developed. • Multivariate Empirical Mode Decomposition (MEMD) to BWR data was applied. • Decision rules permitting to raise an instability alarm based on MEMD. • MEMD can estimate the phase of the density wave from measurements of two LPRMs. • This method detects BWR instabilities (DR increase and out-of phase oscillations). - Abstract: In this work a stability monitor based on a novel technique is presented. This monitor permits to launch general alarms indicating incipient high decay ratios (DR) and out-of-phase oscillations, in a simultaneous way time along. The implemented methodology to determine the estimations of DR and out-of-phase oscillations is based on the Multivariate Empirical Mode Decomposition (MEMD) processing the information obtained from all LPRMs located across the core of Boiling Water Reactor (BWR). The extracted modes with the MEMD, called the Intrinsic Mode Functions (IMFs), permit to tracking the oscillation associated to the density wave. The Case 9 (presenting high DRs and apparently out-of-phase oscillations simultaneously) from the Ringhals stability benchmark was used to show the effectiveness of the proposed methodology

  2. Stability tests in the Grand Gulf unit 1 boiling water reactor

    International Nuclear Information System (INIS)

    This paper summarizes the results of a series of tests performed on January 31, 1987, to determine the stability of the second reload core in the Grand Gulf Unit 1 boiling water reactor (BWR). The subject of BWR stability is relevant for commercial BWR operation. Utilities are required to evaluate reactor stability for every reload core unless plant technical specifications provide for monitoring of neutron flux oscillations in the so-called limit-cycle detect and suppress region at low flows. The parameter of merit for stability calculations or measurements is the asymptotic decay ratio (DR). The definition of asymptotic DR guarantees that as long as its value is < 1.0, the reactor is stable. The DR also yields a quantitative measure of relative stability: DRs below 0.5 are considered very stable. A noise analysis technique was implemented in a portable computer system, which uses standard commercially available hardware, and was used to perform stability measurements on line. This technique has proven to be fairly accurate for high DRs, when the reactor is close to the stability threshold. For low DR conditions, however, the technique yields only reasonable accuracy. An attempt to quantify this accuracy has been made, and the resulting error bands are presented

  3. Verification of advanced methods in TARMS boiling water reactor core management system

    International Nuclear Information System (INIS)

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site Boiling Water Reactor (BWR) core operation management system. It covers almost all the functional requirements to the current process computer to increase on-site core management capability, capacity factors, thermal margins, fuel reliability, and so on, by supporting application functions for monitoring the present core power distribution, and for aiding site engineers in making the core operation plans, by predicting future core performance. It is based on a three dimensional, 1.5 energy group, coarse mesh nodal diffusion theory code ''LOGOS02'', and includes advanced methods to increase the accuracy of core power distribution calculations as well as a local peaking factor calculation method by which the effect of neighboring nodes on intra-nodal power distribution can be considered. TARMS has been installed in eight BWR plants and was verified to be an effective BWR core operation management tool. This paper describes its advanced methods and the results of verifications with actual plant data. (author). 3 refs, 6 figs

  4. Proceedings of the International Workshop on Boiling Water Reactor Stability

    International Nuclear Information System (INIS)

    General design criteria for nuclear power plants in every OECD country require that the reactor core and associated coolant, control, and protection systems be designed so that power oscillations which can result in conditions exceeding acceptable fuel design limits are not possible, or they can be reliably and readily detected and suppressed. In practice, this means that reactor cores should be stable with regard to perturbations from their normal operating state, so that expected variations to the operating parameters do not induce undamped power oscillations. These power oscillations can take a variety of forms, from very local power peaks which can cause no damage, or only slight damage to only a few fuel rods, to large core-wide oscillations where entire segments of the core can become neutronically uncoupled, with wide power swings. Ever since the fast boiling water reactors began operating, over 30 years ago, it has been recognized that their operation under certain conditions of power and flow could cause power and flow oscillations. Considerable research was performed at that time to better understand the principal operating parameters which contribute to the initiation of these oscillations, and guidelines were developed to avoid plant operation under the conditions which were the most unstable. Experiments in the the first Special Power Excursion Reactor Test (SPERT-1) program produced spontaneous power oscillations, and investigations in an out-of-pile loop were necessary to demonstrate that the immediate cause of the oscillations was a power-to-reactivity feedback. Further investigations indicated that the instabilities were limited to certain areas on the operating map. These regions could not be absolutely defined, but there was sufficient understanding of them that they could be generally avoided, with only minor examples of instability events. More recently, though, several reactor events, and especially one that occurred at the La Salle Nuclear

  5. Control method for water quality of BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a method of suppressing radiation exposure upon periodical inspection of a BWR type reactor, suppressing leaching of radioactive materials deposited and activated on fuels, and reducing radioactive deposition on pipelines and equipments made of a carbon steel and austenite stainless steel. Namely, control of water quality described below is conducted under the conditions that the Ni metal ion concentration is from 2 to 10ppb and the Zn metal ion concentration of from 3 to 15ppb in reactor water. (1) controlling the water quality based on neutral/purified water during normal operation and upon injection of hydrogen, (2) using fuels having spring members made of a Ni based alloy processed by aging hardening in atmospheric air, (3) using reactor water recycling pipelines made of an electrolyzed and polished austenite stainless steel, and (4) using carbon steel or low alloy steel for pipelines and equipments of a reactor system. (I.S.)

  6. Mass exchange calculation in a wall layer when water boiling

    International Nuclear Information System (INIS)

    Physical sense and mass exchange characteristics of liquid near-the-wall layer under boiling conditions were attempted to be stated. Equations of material and thermal balance were used to describe the mass exchange characteristics. Technique to calculate circulation ratio in the near-the-wall layer under boiling of under-heated and saturated water was suggested on the basis of the derived expressions. Comparison results of calculated and experimental data were analyzed for full-scale boiling

  7. Boils

    Science.gov (United States)

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

  8. Microbiological Effectiveness of Disinfecting Water by Boiling in Rural Guatemala

    OpenAIRE

    Rosa, Ghislaine; Miller, Laura; Clasen, Thomas

    2010-01-01

    Boiling is the most common means of treating water in the home and the benchmark against which alternative point-of-use water treatment options must be compared. In a 5-week study in rural Guatemala among 45 households who claimed they always or almost always boiled their drinking water, boiling was associated with a 86.2% reduction in geometric mean thermotolerant coliforms (TTC) (N = 206, P < 0.0001). Despite consistent levels of fecal contamination in source water, 71.2% of stored water sa...

  9. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

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

  10. BWR Water Chemistry Guidelines: 1993 Revision, Normal and hydrogen water chemistry

    International Nuclear Information System (INIS)

    The goal of water chemistry control is to extend the operating life of the reactor and rector coolant system, balance-of-plant components, and turbines while simultaneously controlling costs to safeguard the continued economic viability of the nuclear power generation investment. To further this goal an industry committee of chemistry personnel prepared guidelines to identify the benefits, risks, and costs associated with water chemistry in BWRs and to provide a template for an optimized water chemistry program. This document replaces the BWR Normal Water Chemistry Guidelines - 1986 Revision and the BWR Hydrogen Water Chemistry Guidelines -- 1987 Revision. It expands on the previous guidelines documents by covering the economic implications of BWR water chemistry control

  11. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  12. Corrosion Products Identification at Normal Water and Hydrogen Water Chemistry in Boiling Water Reactors

    International Nuclear Information System (INIS)

    The corrosion products sampled from condensate and feedwater systems of boiling water reactors (BWRs) at normal water chemistry (NWC) and hydrogen water chemistry (HWC) operating condition were analyzed with dissolution and instrumental simulation methods. The crystallite and amorphous of iron oxides were separated by means of dissolving method with appropriate chemical solution. The iron oxide composition and content were analyzed by X-ray diffraction (XRD) and inductively coupled plasma atomic emission spectrometer (ICP-AES) in this study. The insoluble iron oxides were obtained in influent and effluent of condensate demineralizer comprised mostly crystalline structure of hematite, magnetite and non-crystallite form of amorphous at NWC and HWC environments. Both goethite and lepidocrocite compositions are of minor importance in feed water system. Crystallite and amorphous compositions in the samples will be calculated from the new developing dissolution method. The crystalline phase of corrosion products are varied with water chemistry conditions in BWRs. The oxide characterization of system corrosion products includes compositions, morphology and particle size can effectively provide the ways of solving crud removal problem in different condition for the performance of condensate demineralizer. The feasibility of identifying other iron oxides and hydroxides in corrosion products is briefly discussed and the mechanisms of iron oxide formation formed around BWR piping will also be shown in detail in this report. Moreover, it will be figured out the properties of radioactive corrosion products growing in different operation periods. The results can also assist in plant units to improve the crud reduction countermeasures and to optimize the system water chemistry. (authors)

  13. EASY 5 BWR simulation model for digital feedwater control design

    International Nuclear Information System (INIS)

    The development of a BWR simulation model in support of a program to design and evaluate the digital feedwater control system for the Monticello Boiling Water Reactor (BWR) is described. This model was developed in the EASY5 simulation language in conjunction with EPRI's Modular Modeling System (MMS) two-phase Library. The model consists of three main elements: the BWR reactor vessel module, the feedwater system model, and the steamline model. Transient results for the BWR vessel module and the feedwater system model are presented

  14. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

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

  15. Predicted impact of power coastdown operations on the water chemistry for two domestic boiling water reactors

    International Nuclear Information System (INIS)

    A theoretical model was adapted to evaluate the impact of power coastdown on the water chemistry of two commercial boiling water reactors (BWRs) in this work. In principle, the power density of a nuclear reactor upon a power level decrease would immediately be lowered, followed by water chemistry variations due to reduced radiolysis of water and extended coolant residence times in the core and near-core regions. It is currently a common practice for commercial BWRs to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal feedwater hydrogen concentration may be different after a power coastdown is implemented in a BWR. A computer code DEMACE was used in the current study to investigate the impact of various power coastdown levels on major radiolytic species concentrations and electrochemical corrosion potential (ECP) behavior of components in the primary coolant circuit of two domestic reactors operating under either normal water chemistry or HWC. Our analyses indicated that under a rated core flow rate the oxidizing species concentrations and the ECP did not vary monotonously with decreases in reactor power level at a fixed feedwater hydrogen concentration. In particular, ECP variations basically followed the patterns of hydrogen peroxide in the select regions and exhibited high values at power levels of 95% and 90% for Chinshan-1 and Kuosheng-1, respectively. (author)

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

    Science.gov (United States)

    Lu, Shanlai

    1997-12-01

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    1991-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

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

    International Nuclear Information System (INIS)

    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

  2. Reducing radiation levels at boiling water reactors of a commercial nuclear power plant fleet

    International Nuclear Information System (INIS)

    Boiling Water Reactors (BWRs) have suffered from high radiation fields in the primary loop, typically measured by the 'BRAC' (BWR Radiation Level Assessment and Control) reactor recirculation system (RRS) dose rates. Reactor water chemistry and activated corrosion product measurements are important in understanding changes in radiation fields in components and systems of a BWR. Several studies have been conducted at Exelon Nuclear's 14 BWRs in order to understand more fully the cause and effect relationships between reactor water radioactive species and radiation levels. Various radiation control strategies are utilized to control and reduce radiation levels. The proper measurement of radioactive soluble and insoluble species is a critical component in understanding radiation fields. Other factors that impact radiation fields include: noble metal applications; hydrogen injection; zinc addition; chemistry results; cobalt source term; fuel design and operation. Chemistry and radiation field trending and projections are important tools that assist in assessing the potential for increased radiation fields and aiding outage planning efforts, including techniques to minimize outage dose. This paper will present the findings from various studies and predictor tools as well as provide recommendations for continued research efforts in this field. Current plant data will be shared on reactor water radioactive species, plant radiation levels, zinc addition amounts and other chemistry controls. (author)

  3. Boiling water reactor shutdown dose rate experience after on-line NobleChem™

    International Nuclear Information System (INIS)

    All U.S. boiling water reactors (BWRs) inject hydrogen for mitigation of intergranular stress corrosion cracking (IGSCC), depleted zinc oxide (DZO) for control of shutdown dose rates, and most have implemented or plan to implement On-Line NobleChem™ (OLNC). In this process, the injection of a platinum compound that catalyzes the recombination of hydrogen and oxygen at surfaces results in restructuring of oxide films on reactor internals and piping, impacting reactor water Co-60 and shutdown dose rates. Since the first implementation of OLNC in 2005, the experience base has significantly expanded in both U.S. and non-U.S. BWRs. This paper investigates the response of reactor recirculation system (RRS) dose rates after OLNC and their relationship to reactor water chemistry parameters, including Co-60 and zinc, using data from EPRI's BWR Chemistry Monitoring and Assessment database. Results of a recent study evaluating correlations of chemistry parameters, other than Co-60, with RRS dose rates are discussed. Relevant revised guidance in the BWR Water Chemistry Guidelines is also presented. (author)

  4. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the 233U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly

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

    Science.gov (United States)

    Thomas, Justin W.

    2006-12-01

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

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

    International Nuclear Information System (INIS)

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

  9. Air–water downscaled experiments and three-dimensional two-phase flow simulations of improved steam separator for boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Katono, Kenichi, E-mail: kenichi.katono.kq@hitachi.com [Hitachi, Ltd., Hitachi Research Laboratory, 1-1, Omika-cho 7-chome, Hitachi-shi, Ibaraki-ken 319-1292 (Japan); Ishida, Naoyuki [Hitachi, Ltd., Hitachi Research Laboratory, 1-1, Omika-cho 7-chome, Hitachi-shi, Ibaraki-ken 319-1292 (Japan); Sumikawa, Takashi; Yasuda, Kenichi [Hitachi-GE Nuclear Energy, Ltd., Hitachi Works, 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken 317-0073 (Japan)

    2014-10-15

    Highlights: • We design the improved steam separator for boiling water reactor (BWR). • The improved steam separator comprises the swirler vanes in the first-barrel section. • We evaluate separator performance by an air–water experiments and flow simulation. • The improved separator can decrease the total pressure losses by about 30%. • And the carryover performance is almost the same level as the conventional separator. - Abstract: Reducing the pressure losses in steam-separator systems in boiling water reactor (BWR) plants is useful for reducing the required pump head and enhancing the design margins to ensure core stability. We need to reduce the pressure losses while maintaining the gas–liquid separation performance. In this study, we improve a steam separator with air–water downscaled experiments and two-phase flow simulations. First, we confirm the effectiveness for the separator performance prediction by adjusting the quality and the two-phase centrifugal force between the air–water downscaled experiments and the steam-water mockup tests, and we design the improved steam separator, which moves the swirl-vane section from diffuser section to the first-barrel section. From the air–water downscaled experiments, the improved separator can decrease pressure loss in the swirler more than 50% around the BWR normal operating conditions compared to the conventional separator, and the carryover of the improved separator is almost the same level as the conventional separator. Next, we evaluate the improved steam separator performance under the BWR operating conditions by means of a two-phase flow simulation, and we have the prospects of the improved separator for reducing the total separator pressure losses by about 30% compared to the conventional separator, while maintaining carryover characteristics.

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

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

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

    International Nuclear Information System (INIS)

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

  12. Potential effects of ex-vessel molten core debris interactions on boiling water reactor containment integrity

    International Nuclear Information System (INIS)

    There is a steadily increasing awareness of the highly plant-specific nature of reactor safety issues. This awareness is reflected in the increasing number of research programs focused on problems limited to specific reactor or containment types. This report is limited to NRC-sponsored research on accident phenomena that may affect the integrity of boiling water reactor containment systems arising out of ex-vessel interactions of molten core debris in the reactor cavity. Some safety issues that are generic to all types of BWRs are discussed, these include: (1) effects of concrete composition, (2) dispersive effect of structures below the reactor vessel, (3) influence of unoxidized zirconium metal in the debris pool, (4) the influence of water in the reactor cavity on debris coolability and magnitude of the radiological source term, and (5) the nature of high-temperature condensed-phase chemistry and fission-product aerosol generation. Certain ex-vessel core-debris phenomena which may threaten the integrity of specific BWR containment designs include the following: (1) integrity of the BWR MARK-I steel pressure boundary, (2) potential for penetration of the MARK-II drywell floor and/or supression-pool bypass, and (3) possible failure of the MARK-III reactor support system due to thermal ablation of the reactor pedestal. Some recent experimental results derived from NRC-sponsored programs are also presented

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

    International Nuclear Information System (INIS)

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

  14. Water chemistry control practices and data of the European BWR fleet

    International Nuclear Information System (INIS)

    Nineteen BWR plants are in operation in Europe, nine built by ASEA Atom, six by Siemens KWU and four by General Electric. This paper gives an overview of water chemistry operation practices and parameters of the European BWR plants. General design characteristics of the plants are described. Chemistry control strategies and underlying water chemistry guidelines are summarized. Chemistry data are presented and discussed with regard to plant design characteristics. The paper is based on a contract of the European BWR Forum with AREVA on a chemistry sourcebook for member plants. The survey of chemistry data was conducted for the years 2002 to 2008. (author)

  15. Improvements in a prototype boiling water reactor: Laguna Verde, Mexico

    International Nuclear Information System (INIS)

    Laguna Verde is the first nuclear power station in Mexico. It has two GE Boiling Water Reactors which will produce 654 MWe each via Mitsubishi turbine generators. At this moment we are ready to load fuel on Unit 1 and 50% complete on Unit 2 beginning electromechanical installation. The project has required 3,600 million dollars including interest rate, over 1,100 full time engineers and about 3,800 direct labour workers and additionally QA, engineering, construction, start-up and operations prepared using approximately 4,400 procedures to perform their activities. Furthermore, 54 industry branches in Mexico have been qualified by quality assurance and they have been providing equipment, components and sub components for the project. Constructing Unit 2 has given us the opportunity to realize the benefits of standardization. Once ''people'' become familiar with a design concept, a BWR-5 with a Mark II containment in this case, the engineering, construction and testing process improves drastically. As of this date, the average savings in man-hours required to build Unit 2 is 40.59% versus the amount needed for Unit 1. We are not making any dramatic change in the design concept of Unit 1, what we are changing in Unit 2 are our working methods and improving when it is appropriate. For instance, large bore piping, HVAC ducts and cable trays are remaining as they are in Unit 1; however, small bore piping, conduit and tubing will be routed in a different manner to reduce as much as possible the number of supports. Supports in Unit 2 will be multidisciplinary since many interferences in Unit 1 were due to an excessive number of supports which were installed on a per discipline basis. We have not achieved that point yet, but in general in control systems, instrumentation and computers there is plenty of room for improvements, by using fiber optics, multiplexers, etc. We will certainly try it. The message is, a developing country does not have the luxury of changing its

  16. Piping reliability analysis for recirculation safe ends of a boiling water reactor

    International Nuclear Information System (INIS)

    This paper presents a piping reliability analysis for the eight recirculation inlet-nozzle safe ends of a boiling water reactor (BWR) nuclear power plant. The analysis is based on principles of probabilistic fracture mechanics. On the basis of observed cracks in the pipe safe ends, the crack is modeled with a semi-elliptical shape initiating at the pipe inner wall. Crack samples are generated using a Monte Carlo simulation technique and an importance sampling scheme. Leak probabilities are estimated through the first ten years of plant lifetime. For the estimated plant operating time of 3 1/2 years, a 20% to 30% probability of safe end leaking is predicted. This prediction correlated well with actual findings at the plant in which one safe end out of eight was leaking after 3 1/2 years. (orig.)

  17. A computational study on instrumentation guide tube failure during a severe accident in boiling water reactors

    International Nuclear Information System (INIS)

    This paper focuses on the nature and timing of Instrumentation Guide Tube (IGT) failure in case of severe core melt accident in a Nordic type Boiling Water Reactor (BWR). First, a 2D structural analysis of a RPV lower head is performed to determine global vessel deformation, timing and mode of failure. Next, a structural analysis is also performed on a 3D IGT section taking into account the influence of global vessel deformation and thermo-mechanical load from the melt pool. We have found that the IG tube was not clamped in the housing at the time when welding ring of the IGT nozzle has been melted and global failure of the vessel wall has not started yet. This suggests that IGT failure is the dominant failure mode in the considered case of a large (~200 tons) melt pool. (author)

  18. Radiation levels at boiling water reactors of a commercial nuclear power plant fleet

    International Nuclear Information System (INIS)

    Radiation field control at a Boiling Water Reactor (BWR) is a complex process that requires the application of both theoretical knowledge and practical experience in order to achieve low radiation fields. Older BWRs were usually designed with cobalt containing components, such as Stellite™ materials in valves, control rod blades, turbine blades and others, that contribute to high radiation fields due to the activation of cobalt to Co-60. Newer BWRs are designed with improvements in these areas; however, only the newest BWRs have been designed using low cobalt source term materials for all components in streams that enter the reactor. Control and minimization of the cobalt source term (material that can be activated to Co-60 in the reactor) will ensure that as low as reasonably achievable (ALARA) dose rates are achieved during power operation and during refueling outages. (author)

  19. A model for fuel rod and tie rod elongations in boiling water reactor fuel bundles

    International Nuclear Information System (INIS)

    A structural model is developed to determine the relative axial displacements of the spring held fuel rods to the tie rods in Boiling Water Reactor fuel bundles. An irradiation dependent relaxation model, which considers a two stage relaxation process dependent upon the fast fluence is used for the compression springs. The changes in spring compression resulting from the change in the length of the zircaloy fuel cladding due to irradiation enhanced anisotropic creep and growth is also considered in determining the time dependent variation of the spring forces. The time dependence of the average linear heat generation rates and their axial distributions is taken into account in determining the fuel cladding temperatures and fast fluxes for the various fuel rod locations within each of the BWR fuel bundles whose relative displacements were measured and used in this verification study. (orig.)

  20. Crack growth of intergranular stress corrosion cracks in austenitic stainless steel pipes of boiling water reactors

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) piping is considered from the crack growth rate point of view. Crack growth rate of sensitized austenitic stainless steel welds is dependent on the degree of sensitization of the material and the severity of the environment as well as the stress state. In evaluation of actual crack growth rate there are three major sources of uncertainty: knowledge of actual crack size and shape, actual stress distribution in he area of the crack and the degree of sensitization. In the report the crack growth calculations used in the USA and in Sweden are presented. Finally, the crack growth rate predictions based on mechanistic modelling of IGSCC and some needs of further research in Finland are considered

  1. Cracking in stabilized austenitic stainless steel piping of German boiling water reactors - characteristic features and root cause

    International Nuclear Information System (INIS)

    Cracks have been found in the welds of piping systems made from stabilized austenitic stainless steels in German boiling water reactors (BWR). In the course of the intensive failure analysis metallographic examinations, microstructural investigations by electron microscopy, corrosion experiments and welding tests have been performed. The results show that cracking under the given medium conditions is due to intergranular stress corrosion cracking (IGSCC) in those parts of the heat affected zone (HAZ) which are overheated during welding and where solution of titanium carbides and subsequent precipitation of chromium carbides and depletion of chromium along the affected grain boundaries could occur. (orig.)

  2. Recent trends in the mitigation of the IGSCC through modifications in the water chemistry of BWR reactors

    International Nuclear Information System (INIS)

    During the last years, the Nuclear Power stations had been that to adequate or to modify the parameters and operational conditions, attempting to maintain and to safeguard the integrity and functionality of its components and systems, as well as the personnel safety involved in its operation. In a Boiling water reactor (BWR), the chemical control of the water, constitutes one of the fundamental aspects to get a sure and reliable operation, having as main objectives: (a) The protection of the reactor vessel, of the structural materials of the same one and of the pipes and components of those recirculation systems against the Intergranular stress corrosion phenomena (IGSCC); (b) To guarantee the integrity of the nuclear fuel minimizing the corrosion phenomena in the fuel elements; and (c) The reduction of the operational dose of the personnel involved directly in the operation and maintenance by means of the control of the activated corrosion products. (Author)

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

    International Nuclear Information System (INIS)

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

  4. Boiling water reactor uranium utilization improvement potential

    International Nuclear Information System (INIS)

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

  5. A diagnostic expert system for a boiling water reactor using a dynamic model

    International Nuclear Information System (INIS)

    A diagnostic expert system for abnormal disturbances in a BWR (Boiling Water Reactor) plant has been developed. The peculiar feature of this system is a diagnostic method which combines artificial intelligence technique with numerical analysis technique. The system has three diagnostic functions, 1) identification of anomaly position (device or sensor), 2) identification of anomaly mode and 3) identification of anomaly cause. Function 1) is implemented as follows. First, a hypothesis about anomaly propagation paths is built up by qualitative reasoning, using knowledge of causal relations among observed signals. Next, the abnormal device or sensor is found by applying model reference method and fuzzy set theory to test the hypothesis, using knowledge of plant structure and function, heuristic strategy of diagnosis and module type dynamic simulator. This simulator is composed of basic transfer function modules. The simulation model for the testing region is built up automatically, according to the requirement from the diagnostic task. Function 2) means identification of dynamic characteristics for an anomaly. It is realized by tuning model parameters so as to reproduce the abnormal signal behavior using the non-linear programing method. Function 3) derives probable anomaly causes from heuristic rules between anomaly mode and cause. A basic plant dynamic model was built up and adjusted to dynamic characteristics for one BWR plant (1100MWe). In order to verify the diagnostic functions of this system, data for several abnormal events was compiled by modifying this model. The diagnostic functions were proved useful, through the simulated abnormal data

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

    International Nuclear Information System (INIS)

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

  7. Feasibility of core management system by data communication for boiling water reactors

    International Nuclear Information System (INIS)

    A core management system by data communication has been designed and proposed for more efficient operation of boiling water reactor (BWR) plants by faster transmission and centralized management of information. The system comprises three kinds f computers: a process computer for monitoring purposes at the reactor site, a center computer for administration purposes at the head office, and a large scientific computer for planning and evaluation purposes. The process and the large computers are connected to the center computer by a data transmission line. To demonstrate the feasibility of such a system, the operating history evaluation system, which is one of the subsystems of the core management system, has been developed along the above concept. Application to the evaluation of the operating history of a commercial BWR shows a great deal of merit. Quick response and a significant manpower reduction can be expected by data communication and minimized intervention of human labor. Visual display is also found to be very useful in understanding the core characteristics

  8. Statics and dynamics of a natural circulation cooled boiling water reactor. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Stekelenburg, A.J.C.

    1994-02-21

    Due to the strong interconnection of the various processes in the reactor vessel of a natural circulation cooled boiling water reactor (BWR), explaining the physics of both the statics and the dynamics of the Dodewaard reactor is not an easy task. In this thesis, the physics is studied through a combined experimental and theoretical investigation. The experiments are analyzed further with the use of the model, and the results of the model calculations provide ideas for new experiments. For an experimental study of the reactor behavior, measurement tools are required. Many relevant process variables are supplied by the power plant's data-logger, but a direct method for measuring the circulation flow rate is not available. Reactor behavior can be studied theoreticallly with the use of a complex computer code, based on a multi-node model. In this way, reliable results are obtained. In many cases, however, such a code is not easy to use, and the calculations require much computer time. Calculations based on a simple model have a lower reliability, but, as the model is clearer, provide more insight into the physics of the system. For this reason, a simple theoretical dynamical model for the main physical processes of the Dodewaard natural circulation cooled BWR is presented in the thesis.

  9. Current status of steam dryer performance under power uprate in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Highlights: • Steam dryer performance after extended power uprate is considered. • Effects of Acoustic Side Branches (ASB) on steam dryers is analyzed. • The ABS represents a reduction in the acoustic loads to the steam dryer. • Spectrograms of signals were obtained for frequency analysis. - Abstract: This work is a compilation of the current status of the steam dryer performance after the implementation of power uprates in Boiling Water Reactors (BWR). Some Nuclear Power Plants (NPPs) have reported failures and cracking in the steam dryer by acoustic resonances that cause excessive vibration due to the increase of steam flow. The replacement of the steam dryer, structural reinforcement and the connection of Acoustic Side Branches (ASB) are the main solutions adopted in order to avoid mechanical failures. The signal analysis of the vibration of the main steam lines in a typical BWR5, was performed using the Short-Time Fourier Transform (STFT). Signals were collected by the strain gauges located around the main steam lines (MSL). Two scenarios are analyzed; the first, is the signal obtained before the installation of the ASB, and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer

  10. Development of open code system for core design of boiling water reactor

    International Nuclear Information System (INIS)

    A new core design system for the Boiling Water Reactor (BWR), HANCS, has been developed. HANCS consists of HIDEC, ALLIS and NORMA, which are open source codes. HIDEC which consists of MVP2.0 and ORIGEN2.1 performs the assembly calculation. ALLIS generates the nuclear constants library for the core calculation. NORMA is introduce in order to perform the core calculation. HANCS was developed by coupling these codes with some other utility programs. HANCS was verified by comparing the calculation results by CASMO-SIMULATE as the reference code. In the verification, the results of the core calculation, such as k-effective, the relative power, the void fraction and the fuel temperature, were compared for the initial loading core and the equilibrium core. In the initial loading core analysis, the calculation results of HANCS agreed well with those of CASMO-SIMULATE under both the zero power condition and the full power operation. In the equilibrium core analysis, although the difference of the void fractions between HANCS and CASMO-SIMULATE was found, the void fractions finally agreed well with those of CASMO-SIMULATE by changing the thermal-hydraulic options of HANCS. The other results also agreed well. It is concluded by the verification that HANCS is appropriate for the BWR core analysis. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-15

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

  13. Investigation on a corrosion product deposit layer on a boiling water reactor fuel cladding

    International Nuclear Information System (INIS)

    Recent investigations on the complex corrosion product deposits on a boiling water reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The magnetic behaviour of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements' oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main phases identified by 2D μXRD mapping inside the layer are hematite and spinel phases with the common formula MxFey(M(1-x)Fe(2-y))O4, where M = Zn, Ni, Mn. It has been shown that the solid solutions of these phases were obtained with rather large differences between the parameter cell of the known spinels (ZnFe2O4, NiFe2O4 and MnFe2O4) and the investigated material. The comparison of EPMA with μXRD analysis shows that the ratio of Fe2O3/MFe2O4 (M = Zn, Ni, Mn) phases in the lower sample equals ∼1/2 and in the higher one ∼1/1 within the analyzed volume of the samples. It has been shown that this ratio, together with the thickness of the corrosion product deposit layer, effect its magnetic properties.

  14. Fracture toughness of highly irradiated stainless steels in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Demma, A. [Electric Power Research Inst., Palo Alto, California (United States); Carter, R. [Electric Power Research Inst., Charlotte, North Carolina (United States); Jenssen, A. [Studsvik Nuclear (Sweden); Torimaru, T. [Nippon Nuclear Fuel Development Co. Ltd, Oarai-machi, Ibaraki (Japan); Gamble, R. [Sartrex Corp., Rockville, Maryland (United States)

    2007-07-01

    Austenitic stainless steels in boiling water reactor (BWR) core structures can experience significant fracture toughness reductions at elevated fluence levels. One of the gaps identified by EPRI is the lack of data over the full range of radiation exposure anticipated for BWRs. This paper describes an experimental project started in 2005 to generate additional fracture toughness data of highly irradiated stainless steels at appropriate fluences, in support of a methodology for evaluating the serviceability of internal components in BWRs. The irradiated austenitic stainless steels retrieved from disposed BWR internal components and their irradiation and fabrication histories are described as well as an updated evaluation of the relationship between fracture toughness and neutron fluence for BWR internals. The effect of specimen orientation on fracture toughness is also being investigated. Microstructural and microchemical analyses of the various materials tested are also presented to complement the fracture toughness results. The fracture toughness results indicate: (1) there is a distinct orientation effect on the toughness, (2) there is no apparent variation in JIC with respect to fluence within the test range (from 3.3 to 9.1 10{sup 21} n/cm{sup 2}, E > 1MeV); any variation with fluence is embedded within the testing and material scatter, and (3) the four specimens corresponding to a material irradiated at approximately 5.2 and 5.9 10{sup 21} n/cm{sup 2} have distinctly lower toughness compared to the other tests. The reason for the low toughness of this material is discussed. (author)

  15. Water injection system for turbine driven BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a water injection system of a turbine driven nuclear reactor for maintaining the function thereof even upon occurrence of a severe accident in a BWR type nuclear reactor. That is, the system comprises a differential pressure detection means for measuring a pressure difference between the downstream of a the turbine and a reactor container and an interrupting means for stopping the supply of steams to the turbine when the differential pressure exceeds a predetermined value. With such a constitution, when the pressure in the turbine driven water injection system is locally increased, the differential pressure detection means detects the differential pressure, to interrupt the supply of the steams to the turbine. Further, upon occurrence of a severe accident that a pressure in the reactor container is abnormally elevated, differential pressure is not caused between the downstream of the turbine and the reactor container. Accordingly, a protection function is not operated by the differential pressure detection means. Accordingly, injection of coolants to the reactor can be continued even upon loss of AC power source. (I.S.)

  16. Eulerian two-phase computational fluid dynamics for boiling water reactor core analysis

    International Nuclear Information System (INIS)

    Traditionally, the analysis of two-phase boiling flows has relied on experimentally-derived correlations. This approach provides accurate predictions of channel-averaged temperatures and void fractions and even peak assembly temperatures within an assembly. However, it lacks the resolution needed to predict the detailed intra-channel distributions of temperature, void fraction and steaming rates that are needed to address the fuel reliability concerns which result from longer refueling cycles and higher burnup fuels, particularly for the prediction of potential fuel pin cladding failures resulting from growth of tenacious crud. As part of the ongoing effort to develop a high-fidelity, full-core, pin-by-pin, fully-coupled neutronic and thermal hydraulic simulation package for reactor core analysis], capabilities for Eulerian-Eulerian two-phase simulation within the commercial Computational Fluid Dynamics code Star-CD are being extended and validated for application to Boiling Water Reactor (BWR) cores. The extension of the existing capability includes the development of wall heat partitioning and bubble growth models, implementation of a topology map based approach that provides the necessary capability to switch between the liquid and vapor as the continuous phase on a cell-by-cell basis and the development of appropriate models for the inter-phase forces that influence the movement of bubbles and droplets. Two applications have been identified as an initial demonstration and validation of the implemented methodology. First, the model is being applied to an Atrium-10 fuel assembly from Cycle 11 of the River Bend Nuclear Power Plant. Second, the model is being applied to an international benchmark problem for validation of BWR assembly analysis methods. (authors)

  17. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    International Nuclear Information System (INIS)

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  18. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.)

  19. On the dynamics of bubbles in boiling water

    International Nuclear Information System (INIS)

    Research highlights: → We devote this work to investigate the bubbles dynamics in boiling water. → A simple experiment of laser scattering was designed to obtain dynamical features. → Correlations and non-exponential distributions were found. → A simple model was able to describe several aspects of the system. - Abstract: We investigate the dynamics of many interacting bubbles in boiling water by using a laser scattering experiment. Specifically, we analyze the temporal variations of a laser intensity signal which passed through a sample of boiling water. Our empirical results indicate that the return interval distribution of the laser signal does not follow an exponential distribution; contrariwise, a heavy-tailed distribution has been found. Additionally, we compare the experimental results with those obtained from a minimalist phenomenological model, finding a good agreement.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-06

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

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

    International Nuclear Information System (INIS)

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

  2. Transition from natural convection or nucleate boiling regime to nucleate boiling or film boiling regime caused by a rapid pressure reduction in highly pressurized and subcooled water

    International Nuclear Information System (INIS)

    Transient boiling processes caused by exponentially decreasing system pressures with various decreasing pressure-reduction periods from the initial heat flux on a horizontal cylinder in a pool of highly subcooled water measured were divided into three groups for low and intermediate initial heat flux in natural convection regime and for high initial heat flux in nucleate boiling. The transitions from low initial heat flux values to stable nucleate boiling occurred independently of the pressure-reduction period values. The transitions from intermediate and high initial heat flux values to stable film boiling occurred for the short pressure-reduction period values, although those to stable nucleate boiling occurred for the long pressure-reduction period values. The mechanism of transient boiling process caused by an exponentially decreasing system pressure with a decreasing pressure-reduction period from an initial heat flux on a horizontal cylinder in a pool of highly subcooled water was clarified on the graph of α/q0.7 versus system pressure with the curves of corresponding fully developed nucleate boiling, incipient nucleate boiling due to unflooded cavities with vapor, and incipient heterogeneous spontaneous nucleation (HSN) due to flooded cavities without vapor. The transitions to stable nucleate boiling from the low initial heat flux values occurred independently of the pressure-reduction period values. The transitions from intermediate and high initial heat flux values in natural convection and nucleate boiling to stable film boiling occurred due to the HSN for short pressure-reduction period values; however those to stable nucleate boiling occurred for long pressure-reduction values. (author)

  3. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

  4. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    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

  5. Interpretation of corrosion potential data from boiling-water reactors under hydrogen water chemistry conditions

    International Nuclear Information System (INIS)

    A method was devised to estimate electrochemical conditions at the entrance to the recirculation piping of a boiling water reactor under hydrogen water chemistry (HWC) conditions from electrochemical corrosion potential (ECP) measurements made in remote autoclaves. The technique makes use of the mixed potential model to estimate ECP in the autoclaves and compares estimates to measured values in an optimization on the concentrations of hydrogen peroxide and oxygen in the recirculation system. The algorithm recognizes that H2O2 decomposes in sampling lines and that transit times between the recirculation system and monitoring points depend upon flow rates and sampling line diameters. An analysis was made of ECP data from three monitoring locations in the Barseback BWR in Sweden, as a function of H2 concentration in the feedwater for two flow rates (5,500 kg/s and 6,300 kg/s for the four recirculation loops). HWC did not displace ECP below a critical value of -0.23 VSHE at the lower flow rate until the reactor water [H2] exceeded 0.15 ppm, corresponding to a feedwater H2 level of > 0.93 ppm. At the higher flow rate of 6,300 kg/s (divided equally between four recirculation loops), protection was not predicted until the feedwater [H2] exceeded 1.2 ppm, corresponding to a reactor water [H2] of ∼ 0.195 ppm. The difference was attributed to the greater persistence of H2O2 at high feedwater [H2] at the higher flow rate, possibly because of the lower transit time from the core to the recirculation system

  6. Two compartment water rod for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matzner, B.; Wolters, B.A.

    1993-07-27

    In a fuel bundle for boiling water nuclear reactor, said fuel bundle is described including a matrix of upstanding fuel rods for undergoing nuclear reaction and generating steam, a lower tie plate for supporting the matrix of fuel rods and admitting liquid water moderator to the fuel bundle from the lower portion of said fuel bundle, an upper tie plate for fastening to at least some of the fuel rods and permitting the outflow of liquid and vapor water moderator from the upper portion of said fuel bundle, a channel surrounding said upper and lower tie plates and said fuel rods therebetween for confining moderator flow between said tie plates and around said fuel rods, and a plurality of vertically spaced apart fuel rod spacers, each said spacer surrounding each said fuel rod at the particular elevation of said spacer for maintaining said fuel rods in side-by-side relation, and a water rod for installation to said fuel bundle for supplying liquid moderator to the upper two phase region of said fuel bundle, the improvement to said water rod comprising: said water rod having a first upper compartment, and a second lower compartment, said upper compartment isolated from said lower compartment; said first upper compartment defining an open, upwardly exposed end for receiving and maintaining water in said upper water rod compartment during the power generating operation of said fuel bundle will fill with liquid by gravity flow from above; means communicated to the bottom portion of said lower compartment for receiving water from said lower portion of said fuel bundle; and, means communicated to the upper portion of said lower compartment for discharging water to the interior of said fuel bundle below the upper most spacer of said fuel bundle whereby discharge to said fuel bundle occurs in said upper two phase region of said bundle.

  7. Two compartment water rod for boiling water reactors

    International Nuclear Information System (INIS)

    In a fuel bundle for boiling water nuclear reactor, said fuel bundle is described including a matrix of upstanding fuel rods for undergoing nuclear reaction and generating steam, a lower tie plate for supporting the matrix of fuel rods and admitting liquid water moderator to the fuel bundle from the lower portion of said fuel bundle, an upper tie plate for fastening to at least some of the fuel rods and permitting the outflow of liquid and vapor water moderator from the upper portion of said fuel bundle, a channel surrounding said upper and lower tie plates and said fuel rods therebetween for confining moderator flow between said tie plates and around said fuel rods, and a plurality of vertically spaced apart fuel rod spacers, each said spacer surrounding each said fuel rod at the particular elevation of said spacer for maintaining said fuel rods in side-by-side relation, and a water rod for installation to said fuel bundle for supplying liquid moderator to the upper two phase region of said fuel bundle, the improvement to said water rod comprising: said water rod having a first upper compartment, and a second lower compartment, said upper compartment isolated from said lower compartment; said first upper compartment defining an open, upwardly exposed end for receiving and maintaining water in said upper water rod compartment during the power generating operation of said fuel bundle will fill with liquid by gravity flow from above; means communicated to the bottom portion of said lower compartment for receiving water from said lower portion of said fuel bundle; and, means communicated to the upper portion of said lower compartment for discharging water to the interior of said fuel bundle below the upper most spacer of said fuel bundle whereby discharge to said fuel bundle occurs in said upper two phase region of said bundle

  8. Electrochemical study of aluminum corrosion in boiling high purity water

    Science.gov (United States)

    Draley, J. E.; Legault, R. A.

    1969-01-01

    Electrochemical study of aluminum corrosion in boiling high-purity water includes an equation relating current and electrochemical potential derived on the basis of a physical model of the corrosion process. The work involved an examination of the cathodic polarization behavior of 1100 aluminum during aqueous oxidation.

  9. Failure analysis of the standby liquid control system for a boiling water reactor with fuzzy cognitive maps

    International Nuclear Information System (INIS)

    Highlights: → FCMs are proposed in order to determine failure modes in systems and equipment in BWRs. → A simplified model is compared with the fault tree analysis technique. → Five case scenarios are studied in order to test the performance of the method. → The proposed method shows consistency with the traditional fault tree technique. - Abstract: A fuzzy cognitive maps (FCM) application is proposed as a simple method to determine failure modes and effects of the standby liquid control system (SLC) during anticipated transient without scram (ATWS) in a boiling water reactor (BWR). The SLC has an important contribution to the total core damage frequency in a BWR. This is the first step in the development of an expert system that could involve many emergency systems of a BWR to simulate accident sequences, through the knowledge representation and reasoning with FCM designs in order to automate the decision making process. A simplified model of the SLC is analyzed with the fault tree analysis technique in order to compare this results with those obtained with the FCM and show consistency with the results, in order to see that both techniques show similar results even if the approaches are different.

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

    International Nuclear Information System (INIS)

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix

  11. Recent trends in the mitigation of the IGSCC through modifications in the water chemistry of BWR reactors; Tendencias recientes en la mitigacion del IGSCC mediante modificaciones en la quimica del agua de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Diaz S, A.; Robles, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    During the last years, the Nuclear Power stations had been that to adequate or to modify the parameters and operational conditions, attempting to maintain and to safeguard the integrity and functionality of its components and systems, as well as the personnel safety involved in its operation. In a Boiling water reactor (BWR), the chemical control of the water, constitutes one of the fundamental aspects to get a sure and reliable operation, having as main objectives: (a) The protection of the reactor vessel, of the structural materials of the same one and of the pipes and components of those recirculation systems against the Intergranular stress corrosion phenomena (IGSCC); (b) To guarantee the integrity of the nuclear fuel minimizing the corrosion phenomena in the fuel elements; and (c) The reduction of the operational dose of the personnel involved directly in the operation and maintenance by means of the control of the activated corrosion products. (Author)

  12. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

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

  13. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas...

  14. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    International Nuclear Information System (INIS)

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  15. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, Duc Ta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-31

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  16. Practical application of neutron noise analysis at boiling water reactors

    International Nuclear Information System (INIS)

    The present status in the development of neutron noise methods for diagnostic purposes at BWRs is assessed with respect to practical applications. Three items of interest are briefly reviewed. They are concerned with local phenomena found in neutron noise signals at the higher frequency ranges (above several Hertz). The detection of vibrating in-core instrument tubes and the impacting of fuel element boxes were a problem in which neutron noise analysis substantially contributed. The possibility of detecting bypass flow boiling from neutron noise signatures is a recently proposed concept. Most of the research efforts have been applied to the experimental determination of local characteristics of the two-phase flow which dominates the noise sources in a BWR. Steam velocity measurements in fuel bundles by neutron noise techniques and the derivation of semi-empirical data, e.g. void fraction, bundle power and inlet flow rate, and possibly flow pattern recognition are features for practical use. But there are still effects which are not yet completely understood and require further experimental and theoretical investigations. (Auth.)

  17. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    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

  18. BWR AXIAL PROFILE

    Energy Technology Data Exchange (ETDEWEB)

    J. Huffer

    2004-09-28

    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.

  19. Heat transfer correlation for saturated flow boiling of water

    International Nuclear Information System (INIS)

    The saturated flow boiling heat transfer of water (H2O, R718) is encountered in many applications such as compact heat exchangers and electronic cooling, for which an accurate correlation of evaporative heat transfer coefficients is necessary. A number of correlations for two-phase flow boiling heat transfer coefficients were proposed. However, their prediction accuracies for H2O are not satisfactory. This work compiles an H2O database of 1055 experimental data points from micro/mini-channels from nine independent studies, evaluates 41 existing correlations to provide a clue for developing a better correlation of saturated flow boiling heat transfer coefficients for H2O, and then proposes a new one. The new correlation incorporates a newly proposed dimensionless number and makes great progress in prediction accuracy. It has a mean absolute deviation of 10.1%, predicting 81.9% of the entire database within ±15% and 91.2% within ±20%, far better than the best existing one. Besides, it also works well for several other working fluids, such as R22, R134a, R410A and NH3 (ammonia, R717), being the best for R22, R410A and NH3 so far. - Highlights: • Compiles a database of 1055 data points of H2O flow boiling heat transfer. • Evaluates 41 correlations of flow boiling heat transfer coefficient. • Generalize approach for developing experiment-based correlation. • Propose a correlation of H2O flow boiling heat transfer in small channels. • The new correlation has a mean absolute deviation of 10.1% for the database

  20. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

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

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

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  2. Age-related degradation of BWR control rod drives

    International Nuclear Information System (INIS)

    This paper reviews the major age-related degradation mechanisms for U. S. boiling water reactor (BWR) control rod drives (CRDs). Component aging caused by various types of stress corrosion cracking, fatigue, general corrosion, wear, and rubber degradation are discussed. (author)

  3. Local heat transfer from the corium melt pool to the boiling water reactor pressure vessel wall

    International Nuclear Information System (INIS)

    The present study considers in-vessel accident progression after core melt relocation to the lower head of a Boiling Water Reactor (BWR) and formation of a melt pool containing a forest of Control Rod Guide Tubes (CRGTs) cooled by purging flows. Descending streams of melt that flow along cooled surfaces of CRGT, and impinge on the bottom surface of the vessel wall can significantly increase local heat transfer. The area of enhanced heat transfer enlarges with decreasing of the melt Prandtl (Pr) number, while the peaking value of the heat transfer coefficient is a non-monotone function of Pr number. The melt Pr number depends on the melt composition (fractions of metallic and oxidic melt components) and thus is inherently uncertain parameter of the core melting and relocation scenarios. The effect of Pr number in the range of 1.02 - 0.03 on the local and integral thermal loads on the vessel wall is examined using Computational Fluid Dynamics (CFD). Heat transfer models obtained on the base of CFD simulations are implemented in the Phase-change Effective Convectivity Model (PECM) for simulation of reactor-scale accident progression heat transfer in real 3D geometry of the BWR lower plenum. We found that the influence of the low Pr number on the thermal loads in a big melt pool becomes more significant at later time, than rapid acceleration of the creep in the vessel wall. This result suggests that global vessel failure is insensitive to the melt composition in the considered 0.7 m deep melt pool configuration. However, it is not clear yet if the low Pr number effect has an influence on vessel failure mode in the other possible melt pool configurations. (author)

  4. A nondiffusive solution method for RETRAN-03 boiling water reactor stability analysis

    International Nuclear Information System (INIS)

    This paper reports that boiling water reactors (BWRs) are susceptible to thermal-hydraulic instabilities that must be considered in BWR design and operation. Early BWRs were designed to be very stable while operating under natural-circulation conditions. As reactor designs have been modified, stability margins have been reduced, and the potential for stability events, such as occurred at the La Salle and Vermont Yankee plants, has increased. These events and other considerations point to the need for a reliable analysis tool for predicting the dynamic behavior of these events. Transient thermal-hydraulic systems analysis codes have been used to analyze hydrodynamic instabilities, and although the results are often reasonable and exhibit the expected behavior, they are sensitive to changes in node and time-step size and a converged solution cannot be demonstrated by reducing the node and time-step sizes. This sensitivity is due to numerical-diffusion that limits the use of most time domain system analysis codes for BWR stability analyses since it directly affects the decay (or growth) ratio compared for stability events. A conservation equation transport model using the method of characteristics has been developed for use with the RETRAN-03 mixture energy and vapor continuity equations. The model eliminates numerical diffusion in the RETRAN solution. The development and validation of a conservation equation transport model for the RETRAN-03 time domain thermal-hydraulic analysis code that extends the range of application to simulating the dynamic behavior of stability events are presented. RETRAN-03 analyses are presented that compare simulations of hydrodynamic instability events with data

  5. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

    A relational database management system with a query language, Reference 1, has been used to develop a Boiling Water Reactor (BWR) fuel channel tracking system on a microcomputer. The software system developed implements channel vendor and Nuclear Regulatory Commission recommendations for in-core channel movements between reactor operating cycles. A BWR Fuel channel encloses the fuel bundle and is typically fabricated using Ziracoly-4. The channel serves three functions: (1) it provides a barrier to separate two parallel flow paths, one inside the fuel assembly and the other in the bypass region outside the fuel assembly and between channels; (2) it guides the control rod as it moves between fuel assemblies and provides a bearing surface for the blades; and (3) it provides rigidity for the fuel bundle. All of these functions are necessary in typical BWR core designs. Fuel channels are not part of typical Pressurized Water Reactor (PWR) core designs

  6. Analyses of Instability Events in the Peach Bottom-2 BWR Using Thermal-Hydraulic and 3D Neutron Kinetic Coupled Codes Technique

    OpenAIRE

    2008-01-01

    Boiling water reactor (BWR) instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presen...

  7. Corrosion product deposition on fuel element surfaces of a boiling water reactor

    International Nuclear Information System (INIS)

    Over the last decade the problem of corrosion products deposition on light water reactor fuel elements has been extensively investigated in relation to the possibility of failures caused by them. The goal of the present study is to understand in a quantitative way the formation of such kind of deposits and to analytically understand the mechanism of formation and deposition with help of the quasi-steady state concentrations of a number of 3d metals in reactor water. Recent investigations on the complex corrosion product deposits on a Boiling Water Reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The buildup of magnetic corrosion product deposits (crud) on the fuel cladding of the BWR, Kernkraftwerk Leibstadt (KKL) Switzerland has hampered the Eddy-current based measurements of ZrO2 layer thickness. The magnetic behavior of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements’ oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main components identified by 2D μXRD mapping inside the layer were hematite and spinel phases with the common formula (MxFey)[M(1-x)Fe(2-y)]O4, where M = Zn, Ni, Mn. With μXRD it was clearly shown that the cell parameter of analyzed spinel is different from the one of the pure endmembers (ZnFe2O4, NiFe2O4 and MnFe2O4) proving the existence of solid solutions. These

  8. 75 FR 26967 - Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water...

    Science.gov (United States)

    2010-05-13

    ... Areas Subject to a Boil-Water Advisory; Availability AGENCY: Food and Drug Administration, HHS. ACTION... entitled ``Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory.'' This guidance is intended to advise food manufacturers that once a boil-water advisory has...

  9. Saturation conditions in elongated single-cavity boiling water targets

    OpenAIRE

    Steyn, G. F.; Vermeulen, C.

    2015-01-01

    Introduction It is shown that a very simple model reproduces the pressure versus beam current characteristics of elongated single-cavity boiling water targets for 18F production surprisingly well. By fitting the model calculations to measured data, values for a single free parameter, namely an overall heat-transfer coefficient, have been extracted for several IBA Nirta H218O targets. IBA recently released details on their new Nirta targets that have a conical shape, which constitutes an im...

  10. Dilute chemical decontamination process for pressurized and boiling water reactor applications

    International Nuclear Information System (INIS)

    Westinghouse Electric Corporation (WEC) has developed five chemical processes for nuclear decontamination, based on extensive experimental testing using radioactive pressurized water reactor (PWR) and boiling water reactor (BWR) samples. The dilute chemical decontamination process offers the best combination of effectiveness, low corrosion, low waste volume, and fast field implementation time. This is an alternating multistep process. For PWRs, an oxidation treatment is necessary. Projected contact decontamination factors (DFs) are about 50 on plant Inconel surfaces, with comparable results on stainless steel. Actual test DFs have exceeded 500 in the process test loop. For BWRs, an oxidation step is unnecessary, but very beneficial. DFs of 10 to 20 are achieved without an oxidation treatment. Full process DFs exceed 500 when the oxidation treatment is included. Low corrosion rates are observed, without any adverse effects. Only solid waste is produced by the process. WEC has fabricated a trailer-mounted application system for this process, and is offering it as a decontamination service to commercial customers

  11. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼ 21 Btu/Ibm. A sensitivity study with regard to the steam separator pressure-loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty in the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼ 500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power

  12. Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

    Energy Technology Data Exchange (ETDEWEB)

    Yan Jin, E-mail: jinyan10@gmail.co [GE-Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, M/L-30, NC 28402 (United States); Bolger, Francis [GE-Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, M/L-30, NC 28402 (United States)

    2010-07-15

    In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.

  13. Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

    International Nuclear Information System (INIS)

    In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.

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

    Science.gov (United States)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

  16. Cycle studies: material balance estimation in the domain of pressurized water and boiling water reactors. Experimental qualification

    International Nuclear Information System (INIS)

    This study is concerned with the physics of the fuel cycle the aim being to develop and make recommendations concerning schemes for calculating the neutronics of light water reactor fuel cycles. A preliminary study carried out using the old fuel cycle calculation scheme APOLLO1- KAFKA and the library SERMA79 has shown that for the compositions of totally dissolved assemblies from Pressurized Water Reactors (type 17*17) and also for the first time, for Boiling Water Reactor assemblies (type 8*8), the differences between calculation and measurement are large and must be reduced. The integration of the APOLLO2 neutronics code into the fuel cycle calculation scheme improves the results because it can model the situation more precisely. A comparison between APOLLO1 and APOLLO2 using the same options, demonstrated the consistency of the two methods for PWR and BWR geometries. Following this comparison, we developed an optimised scheme for PWR applications using the library CEA86 and the code APOLLO2. Depending on whether the information required is the detailed distribution of the composition of the irradiated fuel or the average composition (estimation of the total material balance of the fuel assembly), the physics options recommended are different. We show that the use of APOLLO2 and the library CEA86 improves the results and especially the estimation of the Pu239 content. Concerning the Boiling Water Reactor, we have highlighted the need to treat several axial sections of the fuel assembly (variation of the void-fraction, heterogeneity of composition). A scheme using Sn transport theory, permits one to obtain a better coherence between the consumption of U235, the production of plutonium and burnup, and a better estimation of the material balance. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

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

  18. An in-line diffuse reflection spectroscopy study of the oxidation of stainless steel under boiling water reactor conditions

    International Nuclear Information System (INIS)

    A novel cell unit was constructed to measure in-line the oxide layer build-up on a stainless steel sample by Diffuse Reflection Spectroscopy (DRS; ultraviolet, visible, near infrared) under boiling water reactor (BWR) conditions. The stainless steel samples, observed in the cell through a sapphire window, are contacted with oxidising hot water (300oC, 9.0 MPa). Using a cold finger with the optical fibre probe, the spectroscopic investigations (200-1000 nm wavelength) were performed at a fixed position from the sapphire window. The DRS spectra are the result of the coupling of both absorption (chemical) and interferometric (physical) processes. Analysis of these spectral components allows the independent determination of the oxide layer thickness. The build-up of the oxide layer may be directly observed and quantified, nanometre by nanometre, from 2 to 200 nm. This powerful technique may be used to study the early corrosion rates of stainless steel under BWR conditions and should allow the development of a strategy to reduce corrosion. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki; Arif, Idam; Permana, Sidik [Nuclear Physics and Biophysics Research Division Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

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

  20. Hybrid simulation of boiling water reactor dynamics using a university research reactor

    International Nuclear Information System (INIS)

    A ''hybrid'' reactor/simulation (HRS) testing arrangement has been developed and experimentally verified using The Pennsylvania State University (Penn State) TRIGA Reactor. The HRS uses actual plant components to supply key parameters to a digital simulation (and vice versa). To implement the HRS on the Penn State TRIGA reactor, an experimental or secondary control rod drive mechanism is used to introduce reactivity feedback effects that are characteristic of a boiling water reactor (BWR). The simulation portion of the HRS provides a means for introducing reactivity feedback caused by voiding via a reduced order thermal-hydraulic model. With the model bifurcation parameter set to the critical value, the nonlinearity caused by the neutronic-simulated thermal/hydraulic coupling of the hybrid system is evident upon attaining a limit cycle, thereby verifying that these effects are indeed present. The shape and frequency of oscillation (∼ 0.4 Hz) of the limit cycles obtained with the HRS are similar to those observed in operating commercial BWRs. A control or diagnostic system specifically designed to accommodate (or detect) this type of anomaly can be experimentally verified using the research reactor based HRS

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. TARMS, an on-line boiling water reactor operation management system

    International Nuclear Information System (INIS)

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2015-09-01

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

  6. AXIAL: a system for boiling water reactor fuel assembly axial optimization using genetic algorithms

    International Nuclear Information System (INIS)

    A system named AXIAL is developed based on the genetic algorithms (GA) optimization method, using the 3D steady state simulator code Core-Master-PRESTO (CM-PRESTO) to evaluate the objective function. The feasibility of this methodology is investigated for a typical boiling water reactor (BWR) fuel assembly (FA). The axial location of different fuel compositions is found in order to minimize the FA mean enrichment needed to obtain the cycle length under the safety constraints. Thermal limits are evaluated at the end of cycle using the Haling calculation; the hot excess reactivity and the shutdown margin at the beginning of cycle are also evaluated. The implemented objective function is very flexible and complete, incorporating all the thermal and reactivity limits imposed during fuel design analysis; furthermore, additional constraints can be easily introduced in order to obtain an improved solution. The results show a small improvement in the FA average enrichment obtained with the system related to the reference case that has been studied. The results show that the system converge to an optimal solution, it is observed that the mean fuel enrichment decreases while all the constraints are satisfied. A comparison was also performed using one-point and two-points crossover operator and the results of a sensitivity study for different mutation percentage are also showed

  7. Reactor physics calculations on MOX fuel in boiling water reactors (BWRs)

    International Nuclear Information System (INIS)

    The loading of MOX (Mixed Oxide) fuel in BWRs (Boiling Water Reactors) is considered in this paper in a ''once-through'' strategy. The fuel assemblies are of the General Electric 8 x 8 type, whereas the reactor is of the General Electric BWR/6 type. Comparisons with traditional UOX (Uranium Oxide) fuel assemblies revealed that the loading of MOX fuel in BWRs is possible, but this type of fuel creates new problems that have to be addressed in further detail. The major ones are the SDM (Shutdown Margin) and the stability of the cores at BOC (beginning of cycle), which were demonstrated to be significantly lowered. The former requires a new design of the control rods, whereas a modification of the Pu isotopic vector allows improving the latter. Another issue with the use of the MOX fuel assemblies in a ''once-through'' strategy is the increased radiotoxicity of the discharged fuel assemblies, which is much higher than of the UOX fuel assemblies. (author)

  8. Phased array UT application for boiling water reactor vessel bottom head

    International Nuclear Information System (INIS)

    Stress Corrosion Cracking (SCC) on welds of reactor internals is one of the most important issues in nuclear plants since 1990's. Demands to inspect the reactor internals are increasing. This paper focuses on the development and the application of the phased array ultrasonic testing (PAUT) for the reactor internals located in Boiling Water Reactor (BWR) vessel bottom head (e.g., shroud support). The Toshiba PAUT technologies and technique has been developed and applied to in-Vessel inspections (IVIs) as our universal nondestructive testing (NDT) technologies. Though it was difficult to detect and size cracks in Alloy 182 welds (i.e. weld metal of the shroud support and a CRD stub tube), the efficiency of the PAUT techniques is shown in recent IVI activities. For example the PAUT techniques are applied to crack depth sizing in the weld between the CRD stub tube and RPV bottom build-up in recent years. An immersion technique by the PAUT enables to perform the UT examination on a complex geometric surface to be inspected. The PAUT techniques are developed to detect and size flaws on the shroud support Alloy 182 welds. The techniques include detection from the outside and the inside of RPV. These techniques are applied to the simulated shroud support mockups with SCC-simulated flaws. The examination result is proven to have a good agreement with their actual. As a result, the efficiency of the PAUT techniques is confirmed. (author)

  9. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  10. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  11. The effective convectivity model for simulation of molten metal layer heat transfer in a boiling water reactor lower head

    International Nuclear Information System (INIS)

    The paper is concerned with development of models for assessment of Control Rod Guide Tube (CRGT) cooling efficiency in Severe Accident Management (SAM) for a Boiling Water Reactor (BWR). In case of core melt relocation under a certain accident condition, there is a potential of stratified (with a metal layer atop) melt pool formation in the lower plenum. For simulations of molten metal layer heat transfer we are developing the Effective Convectivity Model (ECM) and Phase-change ECM (PECM). The models are based on the concept of effective convectivity previously developed for simulations of decay-heated melt pool heat transfer. The PECM platform takes into account mushy zone convection heat transfer and compositional convection that enables simulations of non-eutectic binary mixture solidification and melting. The ECM and PECM are validated against various heat transfer experiments for both eutectic and non-eutectic mixtures, and benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is applied to heat transfer simulation of a stratified heterogeneous debris pool in the presence of CRGT cooling. The PECM simulation results show no focusing effect in the metal layer on top of a debris pool formed in the BWR lower plenum and apparent efficacy of the CRGT cooling which can be served as an effective SAM measure to protect the vessel wall from thermal attacks and mitigate the consequences of a severe accident. (author)

  12. Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

    International Nuclear Information System (INIS)

    This paper presents the three-dimensional finite element seismic response analysis of full-scale boiling water reactor BWR5 at Kashiwazaki-Kariwa Nuclear Power Station subjected to the Niigata-ken Chuetsu-Oki earthquake that occurred on 16 July 2007. During the earthquake, the automatic shutdown system of the reactors was activated successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found that there were no significant damages of the reactor cores or other important systems, structures and components through in-depth investigation. In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of structures and components. Although the lumped mass model has worked well so far for a seismic proof design, it is still needed to develop more precise methods for the visual understanding of response behaviors. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize its dynamic behaviors. Through the comparison between both analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value. (author)

  13. Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

    International Nuclear Information System (INIS)

    In this paper, we present the three-dimensional finite element seismic response analysis of the full-scale boiling water reactor BWR5 at the Kashiwazaki-Kariwa Nuclear Power Plant subjected to the Niigata-ken Chuetsu-Oki (NCO) earthquake that occurred on 16th July 2007. During the earthquake, the automatic shutdown of the reactors was performed successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found through in-depth investigation that there was no significant damage of the reactor cores or other important systems, structures and components (SSCs). In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of SSCs. Although the lumped mass model has worked well so far for a seismic proof design, more precise methods should be developed to understand response behaviors visually. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize the dynamic behaviors of this model. Through the comparison of the analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value. (author)

  14. Advanced analytical techniques for boiling water reactor chemistry control

    International Nuclear Information System (INIS)

    The analytical techniques applied can be divided into 5 classes: OFF-LINE (discontinuous, central lab), AT-LINE (discontinuous, analysis near loop), ON-LINE (continuous, analysis in bypass). In all cases pressure and temperature of the water sample are reduced. In a strict sense only IN-LINE (continuous, flow disturbance) and NON-INVASIVE (continuous, no flow disturbance) techniques are suitable for direct process control; - the ultimate goal. An overview of the analytical techniques tested in the pilot loop is given. Apart from process and overall water quality control, standard for BWR operation, the main emphasis is on water impurity characterization (crud particles, hot filtration, organic carbon); on stress corrosion crackling control for materials (corrosion potential, oxygen concentration) and on the characterization of the oxide layer on austenites (impedance spectroscopy, IR-reflection). The above mentioned examples of advanced analytical techniques have the potential of in-line or non-invasive application. They are different stages of development and are described in more detail. 28 refs, 1 fig., 5 tabs

  15. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    Energy Technology Data Exchange (ETDEWEB)

    Rubio, Rafael, E-mail: rrubio@iberdrola.es [Iberdrola Generación Nuclear S.A., Madrid (Spain); Jimenez, Gonzalo [Universidad Politécnica de Madrid (Spain)

    2014-08-15

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.

  16. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    International Nuclear Information System (INIS)

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available

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

    Science.gov (United States)

    Yoshida, Hiroyuki; Takase, Kazuyuki

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

  18. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary

  19. SWR 1000: the Boiling Water Reactor of the future

    International Nuclear Information System (INIS)

    Siemens Power Generation Group (KWU) is currently developing - on behalf of and in close cooperation with the German nuclear utilities and with support from various European partners - Germany's next generation of boiling water reactor. This innovative design concept marks a new era in the successful tradition of boiling water reactor technology and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared lo large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. In addition, a state-of-the-art materials concept featuring erosion-resistant materials and low-cobalt alloys as well as cobalt-free substitute materials ensures a low cumulative dose for operating and maintenance personnel and also minimizes radioactive waste. (author)

  20. SWR 1000: The new boiling water reactor power plant concept

    International Nuclear Information System (INIS)

    Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

  1. In-reactor ECP measurements in BWR plants

    International Nuclear Information System (INIS)

    It has previously been confirmed that the initiation and propagation of Stress Corrosion Cracking (SCC) of stainless steel and Ni-based alloy exposed in the primary water of Boiling Water Reactor (BWR) depend on water chemistry. It is known that the corrosion environment is evaluated with Electrochemical Corrosion Potential (ECP) and SCC susceptibility is high when the ECP is high. Then it also has been confirmed that hydrogen injection in feedwater and Noble Metal Chemical Addition (NMCA) are effective to reduce ECP of reactor components. ECP measurements in BWR plants, which are BWR-3 and BWR-4, were performed in Normal Water Chemistry (NWC), Hydrogen Water Chemistry (HWC) and post-NMCA environments to evaluate the mitigation effect of SCC by HWC and NMCA. ECP measurements were conducted in the lower plenum region, bottom head region, below the core plate and near the bottom of the active fuel region, by installing modified LPRM with ECP electrodes, and at bottom head drain line flange location (BHDL). It is confirmed that the ECP are reduced to less than -200mV(SHE) by 0.9 ppm H2 concentration at feedwater in BWR-3 and 1.1 ppm H2 concentration at feedwater in BWR-4, and that the ECP can be reduced to less than -200mV(SHE) by 0.3 ppm or less H2 concentration after NMCA is applied. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

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

  6. Experimental investigation into the effects of coolant additives on boiling phenomena in pressurized water reactors

    International Nuclear Information System (INIS)

    This study investigates the effects of coolant additives like boric acid on boiling phenomena in pressurized water reactors under conditions as realistic as possible. The effects covered range from subcooled boiling to critical boiling conditions (CHF). The focus of this project lies on flow boiling with up to 40 bar and 250 °C in order to generate a data basis for a possible extrapolation to reactor conditions. The results of the experiments are used to implement and validate new models into CFD-Codes in context to a nationwide German joint research project with the specific aim of improving CFD boiling-models. (author)

  7. Flow boiling of water on nanocoated surfaces in a microchannel

    CERN Document Server

    Phan, Hai Trieu; Marty, Philippe; Colasson, Stéphane; Gavillet, Jérôme

    2010-01-01

    Experiments were performed to study the effects of surface wettability on flow boiling of water at atmospheric pressure. The test channel is a single rectangular channel 0.5 mm high, 5 mm wide and 180 mm long. The mass flux was set at 100 kg/m2 s and the base heat flux varied from 30 to 80 kW/m2. Water enters the test channel under subcooled conditions. The samples are silicone oxide (SiOx), titanium (Ti), diamond-like carbon (DLC) and carbon-doped silicon oxide (SiOC) surfaces with static contact angles of 26{\\deg}, 49{\\deg}, 63{\\deg} and 103{\\deg}, respectively. The results show significant impacts of surface wettability on heat transfer coefficient.

  8. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    International Nuclear Information System (INIS)

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables

  9. Upward Flow Boiling to DI-Water and Cuo Nanofluids Inside the Concentric Annuli

    OpenAIRE

    N. Vaeli; M. M. Sarafraz; Peyghambarzadeh, S. M.; F Hormozi

    2015-01-01

    In this work, flow boiling heat transfer coefficients of deionized water and copper oxide water-based nanofluids at different operating conditions have been experimentally measured and compared. The liquid flowed in an annular space. According to the experiments, two distinguished heat transfer regions with two different mechanisms can be seen namely forced convective and nucleate boiling regions. Results demonstrated that with increasing the applied heat flux, flow boiling heat transfer coef...

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

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico, DF 09340 (Mexico)]. E-mail: gepe@xanum.uam.mx; Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragan 779, Col. Narvarte, Mexico, DF 03020 (Mexico); Vazquez-Rodriguez, Alejandro [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico, DF 09340 (Mexico)

    2006-09-15

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-15

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

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

    International Nuclear Information System (INIS)

    Research highlights: → We present an ant-colony-based system for BWR fuel lattice design and optimization. → Assessment of candidate solutions at 0.0 MWd/kg 235U seems to have a limited scope. → Suitable heuristic rules enable more realistic fuel lattice designs. → The election of the objective has a large impact in CPU time. → ACS enables an important decrease of the initial average U-235 enrichment. - Abstract: This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U235 enrichment and Gd2O3 concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 x 10 BWR fuel lattice with an initial average U235 enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U235 enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U235 enrichment; whereas, the k-infinity was inside the ±100 pcm's range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a powerful tool to tackle this step of

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

    International Nuclear Information System (INIS)

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

  15. Analytical simulation of boiling water reactor pressure suppression pool swell

    International Nuclear Information System (INIS)

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement

  16. Analytical simulation of boiling water reactor pressure suppression pool swell

    Energy Technology Data Exchange (ETDEWEB)

    Widener, S.K.

    1986-01-01

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement.

  17. Operational margin monitoring system for boiling water reactor power plants

    International Nuclear Information System (INIS)

    This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

  18. Crisis behaviour of the reactive recoil of a water jets under conditions of explosive boiling

    International Nuclear Information System (INIS)

    One presents the measurement results of the reactive force of boiling up water jet flowing through short channel into the atmosphere depending on superheating degree and at various evaporation mechanisms. The intensive fluctuation evaporation of water (explosive boiling) and presence of a plane perpendicular to the channel axis are shown to result in abrupt reduction of the reactive recoil value

  19. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact... of Title 10 of the Code of Federal Regulations (10 CFR) for the La Crosse Boiling Water Reactor... modifying or adding EP requirements in Section 50.47, Section 50.54, and Appendix E of 10 CFR part 50 (76...

  20. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI AGENCY...) 73.55, for the LaCrosse Boiling Water Reactor (LACBWR). This Environmental Assessment (EA) has been... revised 10 CFR 73.55 through the issuance of a final rule on March 27, 2009 (74 FR 13926). Section...

  1. Boiling water reactor stability analysis in the time domain

    International Nuclear Information System (INIS)

    Boiling water nuclear reactors may experience density wave instabilities. These instabilities cause the density, and consequently the mass flow rate, to oscillate in the shrouded fuel bundles. This effect causes the nuclear power generation to oscillate due to the tight coupling of flow to power, especially under gravity-driven circulation. In order to predict the amplitude of the power oscillation, a time domain transient analysis tool may be employed. The modeling tool must have sufficient hydrodynamic detail to model natural circulation in two-phase flow as well as the coupled nuclear feedback. TRAC/BF1 is a modeling code with such capabilities. A dynamic system model has been developed for a typical boiling water reactor. Using this tool it has been demonstrated that density waxes may be modeled in this fashion and that their resultant hydrodynamic and nuclear behavior correspond well to simple theory. Several cases have been analyzed using this model, the goal being to determine the coupling between the channel hydrodynamics and the nuclear power. From that study it has been concluded that two-phase friction controls the extent of the oscillation and that the existing conventional methodologies of implementing two-phase friction into analysis codes of this type can lead to significant deviation in results from case to case. It has also been determined that higher dimensional nuclear feedback models reduce the extent of the oscillation. It has also been confirmed from a nonlinear dynamic standpoint that the birth of this oscillation may be described as a Hopf Bifurcation

  2. Boiling water reactor stability revisited: The effects of flashing

    International Nuclear Information System (INIS)

    There are numerous investigations of two-phase flow stability with particular emphasis to BWR stability; these have become increasingly sophisticated and complete over the years. The basic features of a new development and frequency-domain code capable of considering all the channels (bundles) in a BWR, flashing of the coolant at low pressure, full coupling with 3D, two-group neutronics, etc. are described. The basic thermal-hydraulic model is used to study the effects of flashing on stability in a BWR-like channel. The behavior of the channel is highly dynamic. Contrary to what could have been intuitively guessed, the effect of flashing is stabilizing; the reasons and mechanisms leading to this are discussed

  3. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  4. High conversion pressurized water reactor with boiling channels

    International Nuclear Information System (INIS)

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–233U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–233U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm3, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore, some means of

  5. Corrosion products, activity transport and deposition in boiling water reactor recirculation systems

    International Nuclear Information System (INIS)

    The deposition of activated corrosion products in the recirculation loops of Boiling Water Reactors produces increased radiation levels which lead to a corresponding increase in personnel radiation dose during shut down and maintenance. The major part of this dose rate is due to cobalt-60. The following areas are discussed in detail: - the origins of the corrosion products and of cobalt-59 in the reactor feedwaters, - the consolidation of the cobalt in the fuel pin deposits (activation), - the release and transport of cobalt-60, - the build-up of cobalt-60 in the corrosion products in the recirculation loops. Existing models of the build-up of circuit radioactivity are discussed and the operating experiences from selected reactors are summarised. Corrosion chemistry aspects of the cobalt build-up in the primary circuit have already been studied on a broad basis and are continuing to be researched in a number of centers. The crystal chemistry of chromium-nickel steel corrosion products poses a number of yet unanswered questions. There are major loopholes associated with the understanding of activation processes of cobalt deposited on the fuel pins and in the mass transfer of cobalt-60. For these processes, the most important influence stems from factors associated with colloid chemistry. Accumulation of data from different BWRs contributes little to the understanding of the activity build-up. However, there are examples that the problem of activity build-up can be kept under control. Although many details for a quantitative understanding are still missing, the most important correlations are visible. The activity build-up in the BWR recirculation systems cannot be kept low by a single measure. Rather a whole series of measures is necessary, which influences not only cobalt-60 deposition but also plant and operation costs. (author) 26 figs., 13 tabs., 90 refs

  6. Implementation of automated, on-line fatigue monitoring in a boiling water reactor

    International Nuclear Information System (INIS)

    A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to a Japanese operating boiling water reactor (BWR), Tsuruga Unit 1, is described. The system uses the influence function approach and rainflow cycle counting methodology, operates on a workstation computer, and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant-unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computes the fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification number-sign 501. Fatigue values are saved automatically on files at times defined by the user for use at a later time. Of particular note, this paper describes some of the details involved with implementing such a system from the utility perspective. Utility installation details, as well as why such a system was chosen for implementation are presented. Fatigue results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. Although the system is specifically set up to address fatigue duty for the feedwater nozzle location, a generic shell structure was implemented so that any other components could be added at a future time without software modifications. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension

  7. A pilot study for errors of commission for a boiling water reactor using the CESA method

    International Nuclear Information System (INIS)

    Probabilistic Safety Assessment (PSA) typically focuses on the errors leading to the non-performance of required actions (Errors of Omission, EOOs). On the other hand, Errors Of Commission (EOCs) refer to inappropriate, undesired actions that aggravate an accident scenario. The challenges to their treatment in PSA relate to both their identification (which error events should be included in the PSA) and to the quantification of their probabilities. This paper presents the results from a plant-specific study to identify potential EOC vulnerabilities and quantify their risk significance. The study addresses a Boiling Water Reactor (BWR) in Switzerland. It is one of the first EOC analyses ever made for BWRs. The Commission Error Search and Assessment (CESA) method was used to identify EOC scenarios. The EOC probabilities were estimated using the elicitation approach developed as part of the ATHEANA method (A Technique for Human Event Analysis), with input from interviews with plant personnel (with oral as well as written questions). The basis for the quantification was a qualitative analysis of the scenario, the operator response and its procedural basis, and of the opportunities for the EOC and its recovery. The results suggest that the contribution to risk of the most important EOCs is comparable to that of the most important errors of omission, i.e. the required actions typically treated in a PSA; thus, they highlight the significance of EOCs in the overall risk profile of the plant. This study demonstrates the feasibility of a systematic treatment of EOCs for large-scale applications and contributes to understanding the importance of EOCs in the plant risk profile.

  8. Severe accident instrumentation systems for BWR water level and temperature in primary containment vessel measurements

    International Nuclear Information System (INIS)

    The severe accident at TEPCO's Fukushima Daiichi nuclear power station (TF1 accident) in March 2011 brought the lost of the functions of many instrumentation systems. In order to enable the measurements of the important parameters such as reactor water level, temperature and so on even in a case such as the TF1 accident occurs, severe accident instrumentation systems are being developed. In this paper, new system configurations of BWR water level measurement and temperature measurement in primary containment vessels are proposed. Then performance tests for prototype sensors of these measurement systems under high temperature conditions are described. (author)

  9. Progress and current status in water radiolysis calculation for BWR primary circuit

    International Nuclear Information System (INIS)

    In order to assess the material integrity of the BWR primary components from the viewpoint of material degradation, it has been widely recognized that the radiolysis modeling and the corrosion potential modeling are of vital use for the environmental evaluation of the BWR primary systems. At the last Workshop on LWR Coolant Water Radiolysis and Electrochemistry, summary papers on both modeling discussions were presented. The Hydrogen Water Chemistry Task Team of the Atomic Energy Society of Japan is now discussing standard evaluation procedures of hydrogen water chemistry to guarantee mitigating corrosive conditions for core internals in aged BWRs. This time the progress and current status on the procedures are reported. The radiolysis model should include G-values and reaction rate constants as essential parameters and also deal with plant related parameters such as radiation dose rates. A kind of a sensitivity analysis was performed and the key parameters have been identified. It has been rather a consensus that the radiolysis modeling technique has been well acknowledged as a useful tool. The mixed potential model has been widely used to predict the electrochemical corrosion potential as an important index to evaluate the SCC susceptibility of the materials. In this model the corrosion potential is calculated based on the concentrations of the water radiolysis products and other fluid dynamics parameters as flow velocity, equivalent diameter, etc. Since ECPs at all the necessary points in the BWR primary system are not available, the model calculation is regarded as a necessary tool to predict the effectiveness of environmental mitigation measures such as hydrogen water chemistry. Therefore it is necessary to establish a standard procedure for the evaluation of radiolysis product and the ECP. Recently AECL has issued a data base on the rate constants and G-values. The adoption of those data will be discussed and the comparison of the results to the previous ones

  10. Service experience of BWR pressure vessels

    International Nuclear Information System (INIS)

    The overall service experience with Boiling Water Reactor (BWR) pressure vessels has been excellent. The only significant factor that impacted the service performance has been thermal fatigue cracking of feedwater inlet nozzle. This concern has been mitigated by eliminating the source of thermal cycling stress through design and operational changes. Although stress corrosion cracking has occurred in early atypical steam generator vessel designs, analysis and field experience has indicated that this mechanism is not expected in the BWR reactor pressure vessel (RPV). Other limited materials related cracking problems have been associated with RPV stainless steel and nickel-base alloy attachments and penetrations. Solutions to these problems have involved design and materials modifications. Finally, due to the low end of life fluence resulting from the large core-to-RPV-wall water annulus, irradiation embrittlement effects are minimal

  11. The design of large natural circulation BWR's

    International Nuclear Information System (INIS)

    Boiling water reactors (BWR) with natural circulation are applied for capacities up to 60 MWe. Based on scale studies, however, it appears that larger production units are more efficient. It is recommended to investigate the bottlenecks in realizing larger reactors (>1000 MWe). The aim of the study on the title subject is to study to what extent the production capacity of BWRs with natural circulation can be increased. Based on data from the literature a simple analytic method has been chosen and existing BWR designs were compared. Capacities of 1300 MWe appear to be possible. These reactors will have a smaller pin diameter and a lower water supply temperature. Also steam separators with a minor pressure reduction must be available. The reliability of the stability measurement must be increased. Based on the results of this investigation the priorities for research on the design of future BWRs have been determined

  12. Coupled field effects in BWR stability simulations using SIMULATE-3K

    International Nuclear Information System (INIS)

    The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17

  13. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Science.gov (United States)

    2013-06-14

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

  14. Development of ECP models for BWR applications

    International Nuclear Information System (INIS)

    The electrochemical corrosion potential (ECP) of stainless steel has been measured under simulated Boiling Water Reactor (BWR) coolant circuit conditions using a rotating cylinder electrode. Based on the results of measurements an empirical model has been developed to predict the ECP of structure materials in a BVTR primary circuit as a function of H2, O2, and H2O2 concentrations in reactor coolant and water flow velocity. The ECP modeling results using the H2, O2, and H2O2 concentrations calculated by the radiolysis model are compared with the available reactor internal ECP data obtained in an operating reactor

  15. Generation of shockwave and vortex structures at the outflow of a boiling water jet

    Science.gov (United States)

    Alekseev, M. V.; Lezhnin, S. I.; Pribaturin, N. A.; Sorokin, A. L.

    2014-12-01

    Results of numerical simulation for shock waves and generation of vortex structures during unsteady outflow of boiling liquid jet are presented. The features of evolution of shock waves and vortex structures formation during unsteady outflow of boiling water are compared with corresponding structures during unsteady gas outflow.

  16. Orificing of water cross inlet in BWR fuel assembly

    International Nuclear Information System (INIS)

    A nuclear reactor fuel assembly is described comprising a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods, a tubular flow channel member surrounding the fuel rods so as to direct flow of coolant/moderator fluid along the fuel rods, respective upper and lower tie plates at opposite ends of the fuel rods, and a hollow water cross having confronting side walls and a closed lower end wall at an inlet end. The water cross extends centrally through and disposed within the flow channel member so as to provide within the flow channel member separate compartments and to divide the bundle of fuel rods into mini-bundles being disposed in the respective compartments, the water cross including inlet cross flow means formed in the side walls near a lower end of the water cross above the closed end wall and near lower end portions of each of the mini-bundles of fuel rods, which inlet cross flow means provides both selected flow communication into the interior of the water cross and flow communication between the respective mini-bundles for minimizing maldistribution and equalizing flow

  17. BWR 90+ - Nuclear power plant for 21st century

    International Nuclear Information System (INIS)

    BWR 90+ is a boiling water reactor, based on the previous models BWR90 and BWR75, and on the operational experiences gained with six reactors of the previous generation. The development work started in 1994 in co-operation with Teollisuuden Voima Oy (TVO). At present all the boiling water reactor owners participate the cooperation. The objectives of the development were: (1) to develop a boiling water reactor of competitive price level and short construction time, and which meets the latest safety requirements, (2) to itemize the technologies improving the security and competitivity of present plants, and (3) to maintain the expertise of the personnel of the companies participating the development work, and improving the BWR- technology. High power output and short construction time reduce the power generation costs. Large amount of fuel assemblies leads to higher safety margins. Reduction of scram groups from 18 to 16 reduces the amount of components, the assembly space and costs. The reactor technical data is as follows: Thermal power output 4250 MWth; electric power output 1500 MWe, construction time 1500 days, costs 1500 pounds/kWe, no. of fuel assemblies 872, no. of scram groups 16, turbines 1, the capacity factor 90% and the duration of service outage 3 weeks. Specific features of BWR90+ are: short construction time and low costs, risk for connection between wet and dry spaces has been minimized, reactor core remains covered by water during loss-of-coolant accident caused by fuel replacement, Passive collection and cooling of core-melt inside the containment, the containment is not the first wall against the spreading of core-melt, steam explosions and core- concrete interactions have low probabilities, high gas- volume of wet-space reduces the pressure increase during a severe accident, filter-equipped gas removal system forms the final overpressure shield, the containment is cylindrical, and the plant is equipped with digital instrumentation and control

  18. Coupled BWR calculations with the numerical nuclear reactor software system

    International Nuclear Information System (INIS)

    The Numerical Nuclear Reactor (NNR) is a software suite for integrated high-fidelity reactor core simulations including neutronic and thermal-hydraulic feedback. Using solution modules with formulations to reflect the multi-dimensional nature of the system, NNR offers a comprehensive core modeling capability with pin-by-pin representation of fuel assemblies and coolant channels. Originally developed for pressurized water reactors, the NNR analysis capabilities have recently been extended for boiling water reactor (BWR) applications as part of EPRI Fuel Reliability Program. The neutronics methodology is extended to treat non-periodic structure of BWR fuel assemblies, and a new Eulerian two-phase CFD boiling heat transfer model has been integrated with the software system. This paper summarizes the experience with, and results of, the first-of-a-kind coupled calculations as demonstration of a fully-integrated, high-fidelity simulation capability for assessment of margin to crud-induced failure from fuel-duty perspective. (authors)

  19. Application of water jet peening (WJP) technology to BWR core support structures

    International Nuclear Information System (INIS)

    WJP (Water Jet Peening) technology generates the compressive residual stresses on the surface of the material by the extremely high impulse pressure caused by the collapse of cavitation bubbles when high-pressure water jet is discharged in the water. WJP is the promising technology for countermeasure of SCC initiated on the core structures of BWR with some advantages; Collection of foreign materials is not necessary, and WJP is applicable to the complicated structure in narrow space and improves relatively wide area with one scanning. This paper summarizes the effects of WJP on SCC mitigation and describes the development of application technology to BWR core support structures for SCC mitigation. In order to apply to the various welds of core support structures, various kinds of devices have been developed, which can effectively lead the water jet to the welds located in narrow spaces. Prior to the field application, confirmation tests and verification tests were executed on full-scale mockups in order to check the WJP shot parameters and to confirm the generated residual stresses. (author)

  20. Irradiation assisted stress corrosion cracking of stainless alloys in BWR normal water chemistry and hydrogen water chemistry

    International Nuclear Information System (INIS)

    Results are reported from a research project with the objectives to investigate the conditions for irradiation assisted stress corrosion cracking (IASCC), and to explore possible remedies. Constant extension rate tensile (CERT) tests were run on specimens of Type 304 stainless steel (SS), Type 316 SS and Alloy 800, which were irradiated in a commercial BWR to fast neutron fluences ranging from 3·1020 to 4·1021 n/cm2 (E>1 MeV). Some of the specimens were under tensile stress during irradiation. Tests were performed in a test loop using reactor water at a high flow rate. The tests were made with normal BWR water chemistry (NWC), BWR hydrogen water chemistry (HWC) and hydrogen peroxide dope NWC or HWC. In oxidizing environment (NWC or H2O2) doped environments Type 316 SS, and possibly Alloy 800, was much less susceptible to IASCC than the Type 304 SS materials. There was no susceptibility to IASCC below a fluence of 5·1020 n/cm2 (E>1 MeV). At approximately 1·1021 n/cm2 (E>1 MeV) a sharp increase in IASCC susceptibility for Type 304 SS was found. The effects of load during irradiation and flux at constant fluence were also investigated

  1. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼21 Btu/lbm. A sensitivity study with regard to the steam separator pressure - loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty In the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power. Implications of the Present Study: With the exception of guidance to avoid SRV cycling, these preliminary results have provided further support for the validity of this response strategy. SABRE calculations have shown that the reactor is probably slightly unstable in natural circulation operation, but relief valve cycling prevents the occurrence of instabilities at or near design pressure. Thus, unstable operation should not be a concern when boron injection and HPCI are available and depressurization is unnecessary. The reactor water level for injection flow corresponding to HPCI operation has been shown to be acceptable and consistent with earlier estimates based on the NSAC results; however, condensation effects have a significant influence on the equilibrium reactor water level. In addition, the SABRE results reinforce PP and L's concerns that

  2. Multi-cycle boiling water reactor fuel cycle optimization

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, K.; Maldonado, G.I. [University of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

  4. Invited talk on ageing management of boiling water reactors (BWRs)

    International Nuclear Information System (INIS)

    A nuclear power plant is built with a certain design life but by managing the operation of the plant with a well designed in-service inspection, repair and replacement programme of the equipment as required we will be able to extend the operation of the plant well beyond it's design life. This is also economically a paying proposition in view of the astronomical cost of construction of a new plant of equivalent capacity. In view of this, there is a growing trend the world over to study the ageing phenomena, especially in respect of nuclear power plant equipment and system which will contribute towards the continued operation of the nuclear power plants beyond their economic life which is fixed mainly to amortize the investments over a period. Tarapur Atomic Power Station (TAPS) which consists of 2 nos. of Boiling Water Reactor (BWRs) with the presently rated capacity of 160 MWe each has been operating for the past 24 years and is completing its 25th year of service by the year 1994 which was considered as its economic life and the plant depreciation as well as fuel supply agreement were based on this period of 25 years. I will be discussing about the available residual life which is much more than the above (25 years) and the studies we have undertaken in respect of the assessment of this residual life. (author). 2 tabs., 6 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    Lund, A.L.

    1997-11-01

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

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

    International Nuclear Information System (INIS)

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

  7. Experimental investigation on critical heat flux and transition boiling of water flow under increased pressure

    International Nuclear Information System (INIS)

    In connection with reactor safety problems (LOCA) a measuring technique has been developed which enables, within the parameter range of medium pressure (0.11 MPa - 1.20 MPa) and low mass flow densities (10 kg/m2s - 500 kg/m2s), exact experimental investigations of critical heat flux and transition boiling of water under quasi-stationary conditions. The system consists of a vertical, temperature-controlled short test section with water flowing upwards inside; an experimental loop controllable to a large extent; a quick automatic data acquisition, and numeric evaluation procedures. Quasi-stationary measured boiling curves, from nucleate boiling to film boiling (circa 450deg C), demonstrate the importance of pressure, mass flow density, and inlet subcooling, the boiling pressure being the most important parameter. The linear course of the boiling curves during transition boiling is remarkable. A frequently suspected hysteresis of the boiling curve could not be detected. The influence of surface effects (contact angle) clearly decreases with increasing pressure. For the empirical correlation of the measured data by means of indices, a statement was chosen that normalizes the heat flux density of transition boiling to the maximum heat flux density at the beginning of the post-CHF range. As a result, the experimental data obtained, and the correlation developed from them, show a better heat transfer in transition boiling than conservatively assumed in general in literature. The temperature-controlled measurements of complete boiling curves supply data for critical heat flow density and the corresponding wall overheating. A comparison with the uncontrolled operation of the test section shows differences of 5-6% only within the range of measurement accuracy of such experiments. (orig.) With 36 figs., 5 tabs

  8. Modeling and measurement of boiling point elevation during water vaporization from aqueous urea for SCR applications

    Energy Technology Data Exchange (ETDEWEB)

    Dan, Ho Jin; Lee, Joon Sik [Seoul National University, Seoul (Korea, Republic of)

    2016-03-15

    Understanding of water vaporization is the first step to anticipate the conversion process of urea into ammonia in the exhaust stream. As aqueous urea is a mixture and the urea in the mixture acts as a non-volatile solute, its colligative properties should be considered during water vaporization. The elevation of boiling point for urea water solution is measured with respect to urea mole fraction. With the boiling-point elevation relation, a model for water vaporization is proposed underlining the correction of the heat of vaporization of water in the urea water mixture due to the enthalpy of urea dissolution in water. The model is verified by the experiments of water vaporization as well. Finally, the water vaporization model is applied to the water vaporization of aqueous urea droplets. It is shown that urea decomposition can begin before water evaporation finishes due to the boiling-point elevation.

  9. Current status of light water reactor and Hitachi's technical improvements for BWR

    International Nuclear Information System (INIS)

    Gradual technical improvements in Japan over the years has improved the reliability of light water reactors, and has achieved the highest capacity factor level in the world. Commercial operation of Fukushima 2-2 (1,100 MW) of the Tokyo Electric Power Co. was started in February, 1984, as the first standardized BWR base plant, ushering in a new age of domestic light water reactor technology. The ABWR (1,300 MW class) has been developed as Japan's next generation light water reactor, with construction aimed at the latter half of the 1980's. Hitachi's extensive efforts range from key nuclear equipment to various related robots, directed at improving safety, reliability, and the capacity factor, while reducing radiation exposure. This paper presents an outline of Hitachi's participation in the light water reactor's improvement and standardization, and the current status of our role in the international cooperation plan for the ABWR. (author)

  10. Electrochemical potential measurements under simulated BWR water chemistry conditions

    International Nuclear Information System (INIS)

    This paper discusses electrochemical potentials (ECP) measured in a high-temperature test loop under various water chemistry conditions. Several types of reference electrodes were used in this study. Effects of chemical additives were systematically examined, and the shift of ECP was found to be relatively small (±50 mV) by the addition of most common chemicals studied. However, the effects of Cu ions on the ECP of stainless steel and other materials are more significant. The effect of H2O2 was semi-quantitatively determined. The increase of ECP, compared with the dissolved O2 at similar levels was found to range from ∼200 mV to ∼500 mV. The effect is attributed to the oxidizing radicals produced from the catalytic decomposition of H2O2 on the stainless steel surface

  11. An overview of boiling water reactor pipe cracking

    International Nuclear Information System (INIS)

    A review is given of the significant technical progress in solving the intergranular stress corrosion cracking (IGSCC) problem that has been accomplished during the 1976-1980 period. A model for the mechanism of IGSCC was developed and has been used to guide the research and development efforts for providing an engineering solution to the problem. A number of pipe remedies or countermeasures have been developed, qualified and implemented in plants under construction and, to some extent, in operating plants. Other remedies are under development and include the use of hydrogen injection in the feedwater. This remedy has the potential for a practical solution for operating plants. In the area of in-service inspection, a significant advancement in the capability to detect, size and locate IGSCC has been demonstrated with the adaptive learning network system. This system operates automatically with remote control and will significantly reduce the radiation exposure of personnel. In response to the continuing incidents of IGSCC in BWR piping a major four-year research programme has been established. The programme is jointly sponsored by a BWR Owners Group and by the Electric Power Research Institute. The goal of this research programme is to help maintain and improve the availability of BWRs by extending piping service lifetime and by minimising the outage times required for inspection and repair. Non-US BWR utility owners have been invited to participate in the programme. Initial response from these utilities has been favourable. (author)

  12. Examination of overlay repaired BWR pipe joints

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking (IGSCC) in a large number of austenitic stainless steel girth welds in boiling water reactor (BWR) piping has prompted the development of the weld overlay for repair (WOR) as a short-term remedy. It is necessary to examine the deposited overlay weld material for adequate definition of its condition and to monitor the overlaid IGSCC to determine if it grows past the bounds assumed in the design of the repair. This paper reports on NDE techniques evaluated using weld overlaid pipe samples containing known defects, overlaid samples removed from BWR service, and overlaid weld joints in plant. These samples included overlays containing fabrication defects and overlaid pipes containing deep and shallow laboratory- and service-induced IGSCC

  13. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    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)

  14. Electrochemical sensors for application to boiling water reactors

    International Nuclear Information System (INIS)

    An effective measure in combating the intergranular cracking of stainless steel in Boiling Water Reactors (BWRs) is the control of the electrochemical corrosion potential (ECP). It has been found that when the ECP of austenitic stainless steel alloys susceptible to cracking is decreased below -0.230 V(SHE) cracks will not initiate. Similarly, the decrease in potential decreases crack growth rate of existing cracks. Decrease of the potential to acceptable values is accomplished by addition of hydrogen to the reactor feedwater. The amount of hydrogen required is determined by the ECP measured at high temperature either in-situ or from a water sample delivered to an external monitoring station. Both reference and metal sensor electrodes are required to determine the ECP. A multiplicity of reference electrodes are used to verify the validity of the measurements. The reference electrodes, Ag/AgCl, the yttria-stabilized ZrO2 sensor and the platinum electrode are designed for either remote, high radiation environments or accessible monitoring installation at plant operating temperatures. In the former application the support structure for the electrochemical sensor is fabricated from ceramics, usually sapphire, and ceramic-to-metal brazes are used for seals. Metal-to-metal seals are welds. For accessible installations high temperature elastomeric seals are used as long as some periodic maintenance is possible. Just as the reference electrodes are designed for remote or accessible installation, the metal sensor electrodes, principally stainless steel, can be manufactured with ceramic-to-metal brazes or elastomeric seals. The complete electrochemical package, with data acquisition system, is then used by plant personnel to control the feedwater H2 injection rate for environmental crack mitigation

  15. Electrochemical sensors for application to boiling water reactors

    International Nuclear Information System (INIS)

    An effective measure in combating the intergranular cracking of stainless steel in Boiling Water Reactors (BWRs) is the control of the electrochemical corrosion potential (ECP). It has been found that when the ECP of austenitic stainless steel alloys susceptible to cracking is decreased below -0.230 V(SHE) cracks will not initiate. Similarly, the decrease in potential to acceptable values is accomplished by addition of hydrogen to the reactor feedwater. The amount of hydrogen required is determined by the ECP measured at high temperature either in-situ or from a water sample delivered to an external monitoring station. Both reference and metal sensor electrodes are required to determine the ECP. A multiplicity of reference electrodes are used to verify the validity of the measurements. The reference electrodes, Ag/AgCl, the yttria-stabilized ZrO2 sensor and the platinum electrode are designed for either remote, high radiation environments or accessible monitoring installation at plant operating temperatures. In the former application the support structure for the electrochemical sensor is fabricated from ceramics, usually sapphire, and ceramic-to-metal brazes are used for seals. Metal-to-metal seals are welds. For accessible installations high temperature elastomeric seals are used as long as some periodic maintenance is possible. Just as the reference electrodes are designed for remote or accessible installation, the metal sensor electrodes, principally stainless steel, can be manufactured with ceramic-to-metal brazes or elastomeric seals. The complete electro-chemical package, with data acquisition system, is then used by plant personnel to control the feedwater H2 injection rate for environmental crack mitigation

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

    International Nuclear Information System (INIS)

    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)

  17. In-core power sharing and fuel requirement study for a decommissioning Boiling Water Reactor using the linear reactivity model

    International Nuclear Information System (INIS)

    Highlights: • Linear reactivity model (LRM) was modified and applied to Boiling Water Reactor. • The power sharing and fuel requirement study of the last cycle and two cycles before decommissioning was implemented. • The loading pattern design concept for the cycles before decommissioning is carried out. - Abstract: A study of in-core power sharing and fuel requirement for a decommissioning BWR (Boiling Water Reactor) was carried out using the linear reactivity model (LRM). The power sharing of each fuel batch was taken as an independent variable, and the related parameters were set and modified to simulate actual cases. Optimizations of the last cycle and two cycles before decommissioning were both implemented; in the last-one-cycle optimization, a single cycle optimization was carried out with different upper limits of fuel batch power, whereas, in the two-cycle optimization, two cycles were optimized with different cycle lengths, along with two different optimization approaches which are the simultaneous optimization of two cycles (MO) and two successive single-cycle optimizations (SO). The results of the last-one-cycle optimization show that it is better to increase the fresh fuel power and decrease the thrice-burnt fuel power as much as possible. It also shows that relaxing the power limit is good to the fresh fuel requirement which will be reduced under lower power limit. On the other hand, the results of the last-two-cycle (cycle N-1 and N) optimization show that the MO is better than SO, and the power of fresh fuel batch should be decreased in cycle N-1 to save its energy for the next cycle. The results of the single-cycle optimization are found to be the same as that in cycle N of the multi-cycle optimization. Besides that, under the same total energy requirement of two cycles, a long-short distribution of cycle length design can save more fresh fuel

  18. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water...

  19. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  20. Experimental study on a new solar boiling water system with holistic track solar funnel concentrator

    International Nuclear Information System (INIS)

    A new solar boiling water system with conventional vacuum-tube solar collector as primary heater and the holistic solar funnel concentrator as secondary heater had been designed. In this paper, the system was measured out door and its performance was analyzed. The configuration and operation principle of the system are described. Variations of the boiled water yield, the temperature of the stove and the solar irradiance with local time have been measured. Main factors affecting the system performance have been analyzed. The experimental results indicate that the system produced large amount of boiled water. And the performance of the system has been found closely related to the solar radiance. When the solar radiance is above 600 W/m2, the boiled water yield rate of the system has reached 20 kg/h and its total energy efficiency has exceeded 40%.

  1. A novel approach for noble metal deposition on surfaces for IGSCC mitigation of boiling water reactor internals

    International Nuclear Information System (INIS)

    A novel in-situ approach has been developed to deposit noble metals on surfaces of materials commonly used in the nuclear power generating industry. The method involves the injection of a noble metal chemical solution directly into the high temperature water that is in contact with a metal surface to be coated with the noble metal. An effective noble metal coating on a surface can be achieved by maintaining the noble metal concentration at a level of 10 to 100 ppb over a period of 48 hours during the injection process. The surface concentration of the noble metal after the treatment was 2 to 3 atomic %, and the noble metal was present to a depth of 200 to 500 A. The concept of noble metal chemical addition (NMCA) technology was successfully used to create a ''noble metal like'' surface on three of the major nuclear materials, 304 SS, Alloy 600 and Alloy 182. The success of this technology was demonstrated by using constant extension rate tensile (CERT) tests, crack growth rate (CGR) tests and electrochemical corrosion potential (ECP) response tests. The NMCA technology in combination with hydrogen has successfully decreased the ECP of surfaces below the critical cracking potential of -0.230 V(SHE), and prevented both crack initiation and crack propagation in simulated boiling water reactor (BWR) environments

  2. Estimation method of water level behavior in the case of large pressure change in a BWR

    International Nuclear Information System (INIS)

    In a BWR, coolant of core and upper plenum involves so much void volume that free surface level change at downcomer is conspicuous owing to increase and decrease of void volume influenced by pressure change. When mass balance in a reactor vessel becomes non-equilibrium due to steam valve stuck open or feedwater pump trip, difference between liquid level and mixture level becomes very large because of void increased due to depressurization. Therefore, it is very difficult to estimate changes in water level after void exclusion by isolation valve closure etc. So a new parameter ''effective increased void volume'' was contrived to estimate water level in the occurrance of above mentioned phenomena, as a result of consideration about relation between discharged mass and reactor pressure. Degree of water level change under initial operating conditions and reactor pressure change can be estimated by using this parameter. (author)

  3. In-air PIXE for analyzing heavy metals in water boiled in pans

    International Nuclear Information System (INIS)

    The release rates of heavy metals from pans were measured for boiling water as well as for an acidic solution prior to an investigation on the release or sorption of trace elements due to cooking of food by boiling. The boiled samples were condensed and analyzed by means of in-air PIXE. The release of heavy metals was measured for five kinds of pans. For all pans the release rates were considerably more increased by boiling of a 5% solution of acetic acid. Furthermore it was found that by using the alumina coated aluminum pan (alumina pan) the respective release rates of Fe, Cu and Zn were all less than 50 μg per 100 cm2 of the pan surface dipped in the solution, and that monitoring of the contents of aluminum in the boiled solution enabled the estimation of the contribution of metal elements from the pan wall. (orig.)

  4. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    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)

  5. Fracture assessment of a BWR pump nozzle

    International Nuclear Information System (INIS)

    Fracture mechanics calculations are performed to support the non-destructive testing (NDT) qualification programs for pump nozzle investigations of boiling water reactor (BWR) nozzles of reactor pressure vessels (RPVs), with the aim of the determination of qualification defects, which are located in the Inconel 182 weld of the pump nozzle at the bottom of the RPV. The ferritic nozzle and housing have an Inconel buttering and each part is cladded with Inconel 182 before it is mounted. All theses weldments are heat treated after welding; only the connecting weldment between pump housing and nozzle, which is also an Inconel 182 weld, performed on site, is in the as welded condition. (author)

  6. An assessment of boiling as a method of household water treatment in South India.

    Science.gov (United States)

    Juran, Luke; MacDonald, Morgan C

    2014-12-01

    This article scrutinizes the boiling of water in Tamil Nadu and Puducherry, India. Boiling, as it is commonly practiced, improves water quality, but its full potential is not being realized. Thus, the objective is to refine the method in practice, promote acceptability, and foster the scalability of boiling and household water treatment (HWT) writ large. The study is based on bacteriological samples from 300 households and 80 public standposts, 14 focus group discussions (FGDs), and 74 household interviews. Collectively, the data fashion both an empirical and ethnographic understanding of boiling. The rate and efficacy of boiling, barriers to and caveats of its adoption, and recommendations for augmenting its practice are detailed. While boiling is scientifically proven to eliminate bacteria, data demonstrate that pragmatics inhibit their total destruction. Furthermore, data and the literature indicate that a range of cultural, economic, and ancillary health factors challenge the uptake of boiling. Fieldwork and resultant knowledge arrive at strategies for overcoming these impediments. The article concludes with recommendations for selecting, introducing, and scaling up HWT mechanisms. A place-based approach that can be sustained over the long-term is espoused, and prolonged exposure by the interveners coupled with meaningful participation of the target population is essential. PMID:25473989

  7. Critical heat flux of an impinging water jet on a heated surface with boiling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.S. [Andong Institute of Informaion Technology, Andong (Korea); Kim, H.D. [Andong National University, Andong (Korea); Choi, K.W. [Incheon University, Incheon (Korea)

    2000-04-01

    The purpose of this paper is to investigate a critical heat flux(CHF) during forced convective subcooled and saturated boiling in free water jet system impinged on a rectangular heated surface. The surface is supplied with subcooled or saturated water through a rectangular jet. Experimental parameters studied are a width of heated surface, a height of supplementary water and a degree of subcooling. Incipient boiling point is observed in the temperature of 6{approx}8 deg.C of superheat of test specimen. CHF depends on jet velocity for various boiling-involved coolant system. CHF also is proportional to the nozzle exit velocity to the power of n, where n is 0.55 and 0.8 for subcooled and saturated boiling, respectively. CHF is enhanced with a higher jet velocity, higher degree of subcooling and smaller width of a heated surface. (author). 18 refs., 13 figs., 1 tab.

  8. Pressure measurements in boiling particle beds with water at 1 bar

    International Nuclear Information System (INIS)

    Pressures have been measured at the top and bottom of uniformly heated beds of uniform spherical particles with water boiling at atmospheric pressure. Particle sizes used vary from 0.22 to 5 mm diameter and bed heights from 50 to 150 mm. The pressures have been recorded at power levels up to dry-out. The results show how much liquid remains in a boiling bed at different power levels and how the liquid/vapour phase pressure losses vary. The results give a valuable insight into the working of a boiling bed. (author)

  9. Boiling water reactor stability analysis by stochastic transfer function identification

    International Nuclear Information System (INIS)

    The univariate and the bivariate ARMA models are proposed as the stochastic transfer function models for the identification of BWR systems. This technique has been developed as a new method for on-line system identification, optimum control, and malfunction monitoring of nuclear power plants. The relationships between the stochastic transfer function model and the differential equation model are derived. The estimation algorithms are developed through the related covariance functions and Green's function by the least squares method. It has been shown that the stochastic models can also be used for fitting the stochastic data which are contaminated with sinusoidal waves. Both the univariate and the bivariate modeling are applied in the BWR system identification and stability analysis. The univariate modeling is applied to decompose the pressure dynamics from the neutron data. From both of the normal operation data and the perturbation experiment data, the reactor dynamics are consistently estimated. The dynamics of the reactor core are estimated as a second order mode with a natural frequency of 0.4 Hz and a damping ratio of 0.1. The univariate modeling is also applied to monitor the local performance of the coolant channel in the reactor. The transfer functions between system's variables are obtained by use of bivariate modeling. The obtained transfer functions are closely related to the stability analysis of thermal-hydraulics in the reactor. The transition of the system dynamics from normal operation to the perturbation experiment are observed

  10. Development of a fully-consistent reduced order model to study instabilities in boiling water reactors

    International Nuclear Information System (INIS)

    A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to other existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or Cmn*V,D - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate Eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good

  11. Automatic fuel lattice design in a boiling water reactor using a particle swarm optimization algorithm and local search

    International Nuclear Information System (INIS)

    Highlights: ► The automatic procedure was developed to design the radial enrichment and gadolinia (Gd) distribution of fuel lattice. ► The method is based on a particle swarm optimization algorithm and local search. ► The design goal were to achieve the minimum local peaking factor. ► The number of fuel pins with Gd and Gd concentration are fixed to reduce search complexity. ► In this study, three axial sections are design and lattice performance is calculated using CASMO-4. - Abstract: The axial section of fuel assembly in a boiling water reactor (BWR) consists of five or six different distributions; this requires a radial lattice design. In this study, an automatic procedure based on a particle swarm optimization (PSO) algorithm and local search was developed to design the radial enrichment and gadolinia (Gd) distribution of the fuel lattice. The design goals were to achieve the minimum local peaking factor (LPF), and to come as close as possible to the specified target average enrichment and target infinite multiplication factor (k∞), in which the number of fuel pins with Gd and Gd concentration are fixed. In this study, three axial sections are designed, and lattice performance is calculated using CASMO-4. Finally, the neutron cross section library of the designed lattice is established by CMSLINK; the core status during depletion, such as thermal limits, cold shutdown margin and cycle length, are then calculated using SIMULATE-3 in order to confirm that the lattice design satisfies the design requirements.

  12. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  13. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 2: A survey of the accuracy of the Studsvik of America CMS codes

    International Nuclear Information System (INIS)

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. Thus, before performing any kind of calculation with MOx fuels, it is necessary to be able to establish the reliability and the accuracy of these Core Management System (CMS) codes. This report presents a quantitative analysis of the models used in the package. A qualitative presentation is realized in a coming report

  14. Hydraulic performance of pump suction inlets for emergency core cooling systems in boiling water reactors. Containment sump reliability studies. Generic task A-43

    International Nuclear Information System (INIS)

    This document reports on the hydraulic performance of two representative Boiling Water Reactor (BWR) Residual Heat Removal (RHR) suction inlet configurations; namely, those of the Mark I, and Mark II and Mark III designs. Key parameters of interest were air-ingestion levels, vortex types, suction pipe swirl, and the RHR inlet pressure loss coefficient. Tests were conducted with nearly uniform and non-uniform approach flows to the inlets. Flows and submergences were in the range of from 2000 to 12,000 gpm per pipe and 2 to 5 ft, respectively, giving a Froude number range of 0.17 to 1.06. Zero air-withdrawal was measured for both configurations for Froude number equal to or less than 0.8 even under non-unifrom approach flows; likewise, no air-core vortices were observed for the same flow conditions

  15. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shripad T. Revankar; Seungmin Oh

    2003-09-30

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  16. Non linear dynamics of boiling water reactor dynamical system

    International Nuclear Information System (INIS)

    The fifth order phenomenological model of March-Leuba for boiling water reactors include the point reactor kinetics equations for neutron balance and effective delayed neutron precursor groups with one node representation of the heat transfer process and channel thermal hydraulics. This nonlinear mathematical model consists five coupled nonlinear ordinary differential equations. The reactivity feedback (void coefficient of reactivity as well as the fuel temperature coefficient of reactivity), heat transfer process and momentum balance are major reasons for the appearance of nonlinearity in this dynamical system. The linear stability of a dynamical system with the existence of nonlinearity cannot predict a true picture of the stability characteristics of dynamical system; hence nonlinear stability analyses become an essential part to predict the global stable region on the stability map. The linear stable region is analyzed by the eigenvalues. In this stable region all the eigenvalues have negative real parts, but when pair of one of the complex eigenvalues passes transversely through imaginary axis, the dynamical system loses or gain its stability via a Hopf bifurcation and limit cycles emerges from the tip. The study of eigenvalues can predict a few bifurcations. The first Lyapunov coefficient and normal form coefficients can be used for the detection of other bifurcations in the systems. Stable or unstable limit cycles excite from these Hopf points. These limits cycles gains or loses their stability via limit point bifurcation of cycles, period doubling bifurcation of cycles and Neimark-Sacker bifurcation of cycles when one of the parameters of the nuclear dynamical system is varied. The stability of these limit cycles can be studied by Floquet theory and Lyapunov coefficient, but the bifurcations of limit cycles can be investigated only by critical Floquet multiplier which is basically the eigenvalue of the monodromy matrices. The cascade of period doubling

  17. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    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.))

  18. 44-BWR WASTE PACKAGE LOADING CURVE EVALUATION

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing BWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results of 100 percent of the current BWR projected waste stream being able to be disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent upon the referenced waste stream being sufficiently similar to the waste stream received for disposal. (3) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials

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

    International Nuclear Information System (INIS)

    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

  20. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

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

    Science.gov (United States)

    Karve, Atul A.

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

  2. Risk based ISI application to a boiling water reactor

    International Nuclear Information System (INIS)

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

  3. Proceedings of the International Workshop on Boiling Water Reactor Stability

    International Nuclear Information System (INIS)

    With regard to technical understanding of the phenomena, the participants agreed that the causes of instability appear to be well understood, but there are many variables involved, and their correlation with instability conditions is not always certain. Most codes claimed to be capable of predicting oscillations and unstable conditions, based on post-test analyses of data from actual events, but there do not seem to be any blind predictions available which accurately predict an instability event before the actual test results are released. As a result, reactor owners have decided that the best course is to avoid, with sufficient margin, certain regions in the power-flow map where regions of instability are known to exist, rather than try to predict them very accurately. The meeting concluded that the safety significance of BWR instability is rather limited, and current estimates of plant risk do not show it to be a dominant contributor. This is because the installed plant protection systems will shut a reactor down when the oscillations exceed power limits, and any fuel damage which might occur will be localized and containable. However, it was also agreed that an instability event could increase uncertainties in the human error rate, because operators who have never seen an unstable reactor may take actions which are not necessarily the best for the particular situation. In addition, although an instability event may not cause any harm to the public, it may cause some fuel failures, and these are certainly a concern to a reactor owner, for economic and radiation protection reasons. Finally, it was also agreed that BWR instability is certainly considered to be significant by the public, where acceptance of the technology would erode if a plant is perceived to be in an uncontrolled state, regardless of the actual risk inherent in the situation

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

    International Nuclear Information System (INIS)

    A denatured (U-233/Th)O2 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 UO2 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)O2-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O2-fueled BWR should perform similar to a UO2-fueled BWR under all operating conditions. A (Pu/Th)O2-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO2-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

  5. Technology, safety and costs of decommissioning a refernce boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

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

    International Nuclear Information System (INIS)

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

  7. Transient CHF enhancement of saturated pool boiling of water using a honeycomb porous media

    International Nuclear Information System (INIS)

    Several studies have been performed to make clear the transient boiling heat transfer during the exponential heat generation which is occurred in reactivity accident of a nuclear reactor. These researches have been focused on the mechanism of the phenomena mainly, not on the enhancement of the transient boiling heat transfer. In a previous study, we proposed a method of CHF enhancement under steady-state conditions using honeycomb porous plate. The CHF was shown experimentally to be enhanced to more than twice that of a plain surface using honeycomb porous plate. The enhancement is considered to result from the capillary supply of liquid onto the heated surface and the release of generated vapor through the channels. In the present paper, enhancement of the transient critical heat flux in pool boiling by the attachment of a honeycomb-structured porous plate on a heated wire is investigated experimentally using water under saturated boiling conditions. (author)

  8. Analysis of the Boiling Water Reactor Turbine Trip Benchmark with the Codes DYN3D and ATHLET/DYN3D

    International Nuclear Information System (INIS)

    The OECD/NRC Boiling Water Reactor (BWR) Turbine Trip Benchmark was analyzed by the code DYN3D and the coupled code system ATHLET/DYN3D. For the exercise 2 benchmark calculations with given thermal-hydraulic boundary conditions of the core, the analyses were performed with the core model DYN3D. Concerning the modeling of the BWR core in the DYN3D code, several simplifications and their influence on the results were investigated. The standard calculations with DYN3D were performed with 764 coolant channels (one channel per fuel assembly), the assembly discontinuity factors (ADF), and the phase slip model of Molochnikov. Comparisons were performed with the results obtained by calculations with 33 thermal-hydraulic channels, without the ADF and with the slip model of Zuber and Findlay. It is shown that the influence on core-averaged values of the steady state and the transient is small. Considering local parameters, the influence of the ADF or the reduced number of coolant channels is not negligible. For the calculations of exercise 3, the DYN3D model validated during the exercise 2 calculations in combination with the ATHLET system model, developed at Gesellschaft fuer Anlagen- und Reaktorsicherheit for exercise 1, has been used. Calculations were performed for the basic scenario as well as for all specified extreme versions. They were carried out using a modified version of the external coupling of the codes, the 'parallel' coupling. This coupling shows a stable performance at the low time step sizes necessary for an appropriate description of the feedback during the transient. The influence of assumed failures of different relevant safety systems on the plant and the core behavior was investigated in the calculations of the extreme scenarios. The calculations of exercises 2 and 3 contribute to the validation of DYN3D and ATHLET/DYN3D for BWR systems

  9. Reliability innovations for AREVA NP BWR fuel

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to light water reactors worldwide, representing today more than 185,000 fuel assemblies on the world market including more than 63,000 fuel assemblies for boiling water reactors (BWRs). ATRIUM trademark 10 fuel assemblies have been supplied to a total of 32 BWR plants worldwide resulting in an operating experience over 20,250 fuel assemblies. ATRIUM trademark 10XP and ATRIUM trademark 10XM are AREVA NP's most recent fuel assembly designs featuring improved fuel utilization and achieving high margins to operating limits while maintaining very good reliability. Nevertheless, fuel failures are still encountered in all modern and advanced fuel assembly designs leading to significant operating limitations or unplanned shutdowns of nuclear power plants. The majority of fuel failures in BWR plants are caused by debris fretting, with PCI induced failures being a second leading cause. AREVA NP runs programs to study these root causes and to develop product solutions as part of the continuous improvement process within the Zero Tolerance for Failure (ZTF) initiative. The focus of the ZTF initiative is to further upgrade BWR fuel assembly reliability to achieve the goal of failure free fuel. In the following, two major product improvements are described that will significantly contribute to this goal: - Improved FUELGUARD trademark Lower Tie Plate - Chamfered Fuel Pellet Design (orig.)

  10. The control of a boiling water reactor power plant for example Muehleberg

    International Nuclear Information System (INIS)

    Simplified fluid circuit flow diagrams are given for two boiling water reactor types, the first having an outer circuit with the boiling water vessel, turbine, condenser and feed-pump and an inner circuit circulating water within the pressure vessel; the second type has a primary loop for the pressure vessel, a heat exchanger and a secondary loop for the turbine and condenser. The first type has been used at Muehleberg, Leibstadt and Kaiseraugst, and the second at Beznau, and Goesgen. A control circuit illustration is given based on Muehleberg and incorporating a proportional (P) controller in the boiling water side of the outer loop and two PID controllers in the condensate return line. A PI regulator is included in the inner loop. (G.C.)

  11. Damage analysis of ceramic boron absorber materials in boiling water reactors and initial model for an optimum control rod management

    International Nuclear Information System (INIS)

    concept - to calculate the control rod's working life both in the control position and the shut-down position - will automatically lead to an optimization of the control rod strategy. Control rod optimisation is demonstrated by accumulating the total amount of control rods required in a medium-sized BWR up to the total reactor holding period. At least 60% of the first core inventory - for this control rod type an existing EMPIRICAL MODEL is already available - may be used up to the total operating period without any safety loss. Looking to the present disposal situation this concept represents a practical way to reduce all high level waste. In addition benefit of utilizing this concept is that it minimizes tritium emission. Control-rods utilized within Boiling Water Reactors (BWR) are designed for the purpose to control and shape the neutron flux profile in the reactor, to adjust the range of regulation referring to the weight rate of the reactor coolant and thirdly by- shutting down the reactor at any time and under any conditions with regard to nuclear aspects, mechanical integrity and control rod history. The designation control- or shut down rod characterize the particular field of activity for a given control rod. The focal point of my work had shown to be a calculation of the nuclear working life of any control rod design as well as an optimisation method with reference to the holding period for a given control rod inventory as a result of measuring data and a theoretical analysis describing the parameters in a general validity form. (author)

  12. Numerical Analysis of Lead-Bismuth-Water Direct Contact Boiling Heat Transfer

    Science.gov (United States)

    Yamada, Yumi; Takahashi, Minoru

    Direct contact boiling heat transfer of sub-cooled water with lead-bismuth eutectic (Pb-Bi) was investigated for the evaluation of the performance of steam generation in direct contact of feed water with primary Pb-Bi coolant in upper plenum above the core in Pb-Bi-cooled direct contact boiling water fast reactor. An analytical two-fluid model was developed to estimate the heat transfer numerically. Numerical results were compared with experimental ones for verification of the model. The overall volumetric heat transfer coefficient was calculated from heat exchange rate in the chimney. It was confirmed that the calculated results agreed well with the experimental result.

  13. Unsteady thermodynamic nonequilibrium wave outflow of boiling water from vessels with branch pipes

    International Nuclear Information System (INIS)

    Within the framework of one-speed thermodynamic nonequilibrium model of bubble structure steam-water mixture flow the numerical solution of the conservation equations system combined with the kinetic equation describing the volume water boiling up on impurity particles in relaxation approximation is performed. The nonstationary wave outflow of boiling water from straight pipes and vessels with branch pipes is investigated. It is shown that the intense evaporation process behing the front of the rarefaction wave arising in the channel in the curse of depressurization is preceded the medium pressure region (elastic precursor)

  14. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    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)

  15. 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)

  16. An A BWR demonstration simulator for training and developing technical staff

    International Nuclear Information System (INIS)

    The US-Advanced Boiling Water Reactor (A BWR), certified by the 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. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  17. An A BWR demonstration simulator for training and developing technical staff

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the 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. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  18. Pool boiling of water on nano-structured micro wires at sub-atmospheric conditions

    Science.gov (United States)

    Arya, Mahendra; Khandekar, Sameer; Pratap, Dheeraj; Ramakrishna, S. Anantha

    2015-10-01

    Past decades have seen active research in enhancement of boiling heat transfer by surface modifications. Favorable surface modifications are expected to enhance boiling efficiency. Several interrelated mechanisms such as capillarity, surface energy alteration, wettability, cavity geometry, wetting transitions, geometrical features of surface morphology, etc., are responsible for change in the boiling behavior of modified surfaces. Not much work is available on pool boiling at low pressures on microscale/nanoscale geometries; low pressure boiling is attractive in many applications wherein low operating temperatures are desired for a particular working fluid. In this background, an experimental setup was designed and developed to investigate the pool boiling performance of water on (a) plain aluminum micro wire (99.999 % pure) and, (b) nano-porous alumina structured aluminum micro wire, both having diameter of 250 µm, under sub-atmospheric pressure. Nano-structuring on the plain wire surface was achieved via anodization. Two samples, A and B of anodized wires, differing by the degree of anodization were tested. The heater length scale (wire diameter) was much smaller than the capillary length scale. Pool boiling characteristics of water were investigated at three different sub-atmospheric pressures of 73, 123 and 199 mbar (corresponding to T sat = 40, 50 and 60 °C). First, the boiling characteristics of plain wire were measured. It was noticed that at sub-atmospheric pressures, boiling heat transfer performance for plain wire was quite low due to the increased bubble sizes and low nucleation site density. Subsequently, boiling performance of nano-structured wires (both Sample A and Sample B) was compared with plain wire and it was noted that boiling heat transfer for the former was considerably enhanced as compared to the plain wire. This enhancement is attributed to increased nucleation site density, change in wettability and possibly due to enhanced pore scale

  19. Synergistic effect of zinc injection and noble metal treatment on the IGSCC, and the improvement of hydrogen water chemistry management technique in BWR coolant

    International Nuclear Information System (INIS)

    Control effects of stress corrosion cracking (SCC) using the water treatment techniques were evaluated by experiments. SCC sensitivity was evaluated by the short strain rate tensile (SSRT) method. Zinc injection did not affect on the control of SCC. With using both zinc injection and noble metal treatment, zinc and noble metals were contained in the surface of stainless steel, but zinc did not affect on the SCC sensitivity of stainless steel. The phenomena that good SCC control electric potential of stainless steel is less than about -0.23V at 288degC was explained theoretically on the basis of transforming NiFe2O4 to FeCr2O4 in the grain boundary of surface oxide depend on hydrogen injection. Under the ordinary running conditions, NO3- is the chemical species of nitrogen in the aqueous solution of the boiling water reactor (BWR). However, chemical formula of nitrogen transformed NO3- → NO2- → NH3 (aq.) by hydrogen injection. To control SCC and decrease dose rate, the electric potential should be decrease by hydrogen injection until NO3- was no longer detectable. (S.Y.)

  20. Cathodic polarization curves of the oxygen reduction reaction on various structural materials of boiling water reactors in high temperature-high purity water

    International Nuclear Information System (INIS)

    Cathodic polarization curves of the O2 reduction reaction were measured by using electrodes made from typical structural materials of boiling water reactors (BWRs) to evaluate the effects of kind of material on the electrochemical corrosion potential (ECP) calculation. To estimate ECPs at any region in the BWRs on the basis of the BWR environmental conditions, anodic and cathodic polarization curves should be obtained in advance under relevant conditions. The concentration of oxidants such as O2 and H2O2 in coolant changes depending on the region in which they exist. As well, reduction reaction rates might differ depending on the kind of materials. In this work, the cathodic polarization curves of type 316L stainless steel (316L SS) and Alloy 182 were measured in high purity water at 553 K with different O2 concentrations and compared with those of type 304 SS (304 SS). The results showed that the cathodic polarization curves differed depending on the kind of materials at the activation-controlled region. But, the difference in the ECP vs. O2 concentration relationship was small when the ECPs were calculated by using both anodic and cathodic polarization curves measured on the objective material. (author)

  1. Identification of chromium oxides and other solids in BWR reactor water

    International Nuclear Information System (INIS)

    Radioactive solid particles in reactor water may deposit as hot spots on reactor component surfaces, contributing to plant radiation field build-up. Phase identification of these solid particles would improve our understanding about the origins of the 'hot spots' and their behaviour under various water chemistry conditions. Phase identification is also important for the purpose of experimental verification of some thermodynamic calculations that predict thermodynamic stability of certain solid phases in BWR water environments. This paper concerns a transmission electron microscopy study on solid particles that were collected from two Swedish BWRs operated with hydrogen water chemistry. In the samples collected from both reactors, a significant fraction of the total activities came from radionuclide Cr-51. Among various solid particles detected, a significant number of chromium oxide particles were found. From one reactor amorphous chromium oxide particles were detected while from another reactor crystalline Cr2O3 was found. The presence of the metastable amorphous chromium oxide in the coolant suggests that any assumption of achieving thermodynamic equilibrium in the coolant system would not be valid. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

  4. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then 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. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. 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 vertical, canistered dry cask systems. 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 deg 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

  5. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    International Nuclear Information System (INIS)

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then 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. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. 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 vertical, canistered dry cask systems. 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 deg 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

  6. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  7. Generic safety evaluation report regarding integrity of BWR scram system piping

    International Nuclear Information System (INIS)

    Safety concerns associated with postulated pipe breaks in the boiling water reactor (BWR) scram system were identified during the staff's continuing investigation of the Browns Ferry Unit 3 control rod partial insertion failure on June 28, 1980. This report includes an evaluation of the licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service. A discussion of the means available for mitigation an unlikely SDV system failure is provided. Generic recommendations are made to improve mitigation capability and ensure that system integrity is maintained in service

  8. Zircaloy spacer grid for boiling light water reactors

    International Nuclear Information System (INIS)

    The need to increase the neutronic efficiency of the new cores of BWR's, lead to study types of spacer-grids made of low neutronic absorption materials as zircaloy-4. The particular mechanical behaviour of this material suggested to design a spacer-grids such as to utilize only blanking, slotting and bending operations as plastic forming and to avoid therefore drawing effects. The optimization of the bending procedures lead to a final spacer-grids configuration equally stiff in all directions and planes. Only for the ''elastic constraints'' nichel alloy sheets were used to made easy the whole spacer design. The ''rigid constraints'', supporting the rods, have been obtained directly from the spacer structure. Calculations were performed to verify the mechanical strength of the main grid components. In this framework a computer code was developed to find the best elastic characteristic of the ''elastic constraints'' taking into account the machining tolerances. Some original methods to test the integral behaviour of the grid assembled as well as the procedures to be adopted for its best maintenance, are described

  9. Radiation effects in organic paints of a Boiling Water Reactor

    International Nuclear Information System (INIS)

    The coatings on a BWR are used as a protection for the building and equipments from corrosion and contamination by radionuclides. The purpose of this work is to test this kind of coatings by simulating real absorbed doses in 40 years of use plus a nuclear accident (LOCA). Standards said that irradiation should be made with gamma radiation. In this work it's suggested to irradiate with electrons simulating secondary radiation produced on the interaction gamma-matter, and protons simulating the damage caused by the interaction neutron-matter. It's also suggested a new kind of adhesion test for coatings that gives a quantitative measure all other tests are qualitative. Two types of coatings were tested: Modified Phenolic and Epoxic both kinds had a very satisfactory performance in all the tests. The maximum dose accumulated by the coatings was 450 Mrad and the minimum 50 Mrad. The dose rates were: gamma in between 0.4 Mrad/hr and 1.0 Mrad/hr; protons and electrons between 500 Mrad/hr and 4000 Mrad/hr. Other important fact is that a calibration was made for a polymer to be used as a high dose dosimeter, these new dosimeters can measure doses between 10 Mrad and 100 Mrad not depending on the dose rate. (author)

  10. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  11. Water jet intrusion into hot melt concomitant with direct-contact boiling of water

    International Nuclear Information System (INIS)

    Boiling of water poured on surface of high-temperature melt (molten metal or metal oxide) provides an efficient means for heat exchange or cooling of melt. The heat transfer surface area can be extended by forcing water into melt. Objectives of the present study are to elucidate key factors of the thermal and hydrodynamic interactions for the water jet injection into melt (Coolant Injection mode). Proposed applications include in in-vessel heat exchangers for liquid metal reactor and emergency measures for cooling of molten core debris in severe accidents of light water reactor. Water penetration into melt may occurs also as a result of fuel-coolant interaction (FCI) in modes other than CI, it is anticipated that the present study contributes to understand the fundamental mechanism of the FCI process. The previous works have been limited on understanding the melt-water interaction phenomena in the water-injection mode because of difficulty in experimental measurement where boiling occurs in opaque invisible hot melt unlike the melt-injection mode. We conducted visualization and measurement of melt-water-vapor multiphase flow phenomena by using a high-frame-rate neutron radiography technique and newly-developed probes. Although limited knowledge, however, has been gained even such an approach, the experimental data were analyzed deeply by comparing with the knowledge obtained from relevant matters. As a result, we succeeded in revealing several key phenomena and validity in the conditions under which stable heat transfer is established. Moreover, a non-intrusive technique for measurement of the velocity and pressure fields adjacent to a moving free surface is developed. The technique is based on the measurement of fluid surface profile, which is useful for elucidation of flow mechanism accompanied by a free surface like the present phenomena. (author)

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

    International Nuclear Information System (INIS)

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

  13. Application of water jet penning technology to BWR core shroud for IGSCC mitigation

    International Nuclear Information System (INIS)

    Water Jet Peening (WJP) is one of the promising SCC mitigation technologies which make original surface tensile residual stress to compressive one. The Water Jet Peening Technology has the following advantages: a) no foreign material entering into the reactor because of using only water, b) applicability to narrow and complicated structure because it is effective in the wide range of parameters, c) simple in the system/equipment and short period of application in actual plant. WJP was first applied to BWR Core Shroud for preventive maintenance purpose during 1999 outage in Japan. Although the target welds of Shroud are surrounded by various kinds of other components and access space is very limited, most of the weld could be peened by optimizing the peening condition. Effect of residual stress improvement was verified by mock-up test prior to actual work. WJP application was completed within the planned schedule without trouble. Application experience to the Shroud and examples of development of application to other Reactor Internal components will be presented. (author)

  14. Augmentation of forced flow boiling heat transfer by introducing air flow into subcooled water flow

    International Nuclear Information System (INIS)

    The effect of air injection into a subcooled water flow on boiling heat transfer and a critical heat flux (CHF) was examined experimentally. Experiments were conducted in the range of subcooling of 50 K, a superficial velocity of water and air Ul = 0.17 ∼ 3.4 and Ug = 0 ∼ 15 m/s, respectively. A test heat transfer surface was a 5 mm wide, 40 mm long and 0.5 mm thick stainless steel sheet embedded on the bottom wall of a 10 mm high and 20 mm wide rectangular flow channel. Nine times enhancement of the heat transfer coefficient in the non-boiling region was attained at the most by introducing an air flow into a water single-phase flow. The heat transfer improvement was prominent when the water flow rate was low and the air introduction was large. The present results of the non-boiling heat transfer were well correlated with the Lockhart-Martinelli parameter Xtt; hTP/hL0 = 5.0(1/ Xtt)0.5. The air introduction has some effect on the augmentation of heat transfer in the boiling region, however, the two-phase flow effect was little and the boiling was dominant in the fully developed boiling region. The CHF was improved a little by the air introduction in the high water flow region. However, that was rather greatly reduced in the low flow region. Even so, the general trend by the air introduction was that qCHF increased as the air introduction was increased. The heat transfer augmentation in the non-boiling region was attained by less power increase than that in the case that only the water flow rate was increased. From the aspect of the power consumption and the heat transfer enhancement, the small air introduction in the low water flow rate region seemed more profitable, although the air introduction in the high water flow rate region and also the large air introduction were still effective in the augmentation of the heat transfer in the non-boiling region. (author)

  15. Analytical and experimental simulation of boiling oscillations in sodium with a low-pressure water system

    International Nuclear Information System (INIS)

    An experimental and analytical program designed to simulate sodium boiling under low-power, low-flow conditions has been completed. Experiments were performed using atmospheric- pressure water as a simulant fluid and a simple one-dimensional model was developed for the system. Results indicate that water is a suitable simulant for liquid sodium under certain conditions and that the model does a fair job of modeling the system. In addition, oscillations that occur during the boiling process appear to augment substantially the heat transfer between liquid and vapor in condensation

  16. Coherent Calculation for Air-Water Flow and Boiling Flow by Using CUPID Code

    International Nuclear Information System (INIS)

    The Korea Atomic Energy Research Institute has been developing a three-dimensional thermal-hydraulic code, called CUPID, which was motivated from practical needs for the realistic simulation of two-phase flows in nuclear reactor components. This paper presents coherent simulation of an air-water flow test and a sub-cooled boiling flow test, and the model implementation of related to them. The closure relations for the air-water flow and sub-cooled boiling flow are turbulence model, interfacial non-drag force, interfacial condensation, wall evaporation model, interfacial area transport equation, and so on

  17. Propagation of cracks by stress corrosion in conditions of BWR type reactor

    International Nuclear Information System (INIS)

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  18. Equilibrium chemistry of nitrogen and potential-pH diagrams for the Fe-Cr-H2O system in BWR water

    International Nuclear Information System (INIS)

    A thermodynamic analysis of the nitrogen-water system for BWR operating conditions assesses the chemical species responsible for steam line radiation increases. Nitrogen (N2) was the important gaseous form during hydrogen injection into BWR feedwater. The potential-pH diagram for the system Fe-Cr-H2O was derived from thermodynamic information for BWR operating temperatures. The diagram shows that a mixed oxide (FeCr2O4) is stable under the reducing conditions imposed by hydrogen water chemistry (HWC) for BWR while the simple oxide Fe2O3 is stable for the more oxidizing conditions of normal water chemistry. The presence of Cr in the oxide may account for the mitigation of intergranular stress corrosion cracking of sensitized stainless steel during HWC. (orig.)

  19. Heat Transfer From Electrically Heated Nichrome Wires to Boiling Water at Different Pressures

    Directory of Open Access Journals (Sweden)

    Devi Dayal

    1968-01-01

    Full Text Available Boiling curves for nucleate and film boiling have been drawn for nichrome of three sizes in distilled and degasified water at saturation temperatures under five different sub-atmospheric vapour pressure. It has been observed that (i for the same Q/A (heat transfer, Delta Theta (excess of wire temperature over saturation point of water decreases with pressure in both nucleate and film boiling ranges, (ii Both Q/A max. and Delta Theta/SubC show a rapid decrease with pressure but these variations become more gradual at higher pressures, and (iii Q/A max. and Delta Theta/SubC increase with wire size at all pressures; increase in Delta Theta/SubC however, becomes less conspicuous at higher pressures approaching one atmosphere.

  20. On the determination of boiling water reactor characteristics by noise analysis

    International Nuclear Information System (INIS)

    In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

  1. Water radiolysis effect on IASCC growth behavior in BWR water conditions in highly irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    For study of water radiolysis effect caused by gamma-rays from radioactive material on irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth tests in highly irradiated austenitic stainless steel are performed in simulated BWR water conditions (at ∼563 K). The compact tension (CT) specimens made of 316L stainless steels are irradiated with neutrons up to ∼12 dpa in the Japan Materials Testing Reactor (JMTR). Post-irradiation annealing (PIA) at 973 K for 1 hour is applied to one of the specimens, which shows the recovery of material properties corresponding to the unirradiated ones but the radioactivity of highly irradiated material as it is. The gamma-ray absorbed dose rate in water is calculated near the crack tip of the CT specimen, and the stable concentrations of H2O2, O2 and H2 in water near the crack tip are estimated by radiolysis calculation for some feed water conditions of normal water chemistry (NWC), deaerated water and hydrogen water chemistry (HWC). The preliminary results of the crack growth rate (CGR) for the highly irradiated specimens and the annealed specimen are presented, and the relationship between the CGRs and the water chemistry such as the concentrations of radiolytic species and the electrochemical corrosion potential (ECP) is discussed. (author)

  2. Influence of metal addition to BWR water on contamination and corrosion of stainless steel

    International Nuclear Information System (INIS)

    Oxide layers grown on stainless steel under modified BWR conditions with or without addition of different bivalent metal ions have been characterised using methods like SIMS and photo-electrochemistry. The Co-58 activity of the samples depends strongly on the thickness of the oxide film. Low pHT values generally favour dominance of p-type semiconductivity, implying a corrosion process controlled by cation transport through the oxide layer. High pHt values normally result in a change of semiconducting properties from predominantly p-type to n-type, which can be used as an indicator for a change in the corrosion mechanism. The metal ions added to high-temperature water do not significantly affect the specific activity of the different stainless steel samples after exposure. The aim of the described tests was to identify possible alternatives to zinc and to elucidate underlying mechanisms controlling the incorporation of radio-isotopes of cobalt in the oxide layer on stainless steel. Manganese has been identified in the described short-term exposure tests as a possible alternative to zinc. The positive effect of manganese must be confirmed by long-term tests. During these tests, the concentration of dissolved manganese should be reduced to 10 ppb maximum and the other water chemistry parameters should be also adjusted more closely to reactor coolant conditions. (orig./MM)

  3. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE-ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  4. Tritium in liquid phase in a BWR-5 like Laguna Verde; Tritio en fase liquida en un BWR-5 como Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Vargas A, A.; Cardenas J, J., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Veracruz-Medellin Km 7.5, Veracruz (Mexico)

    2011-11-15

    In boiling water reactors (BWR), the tritium (H{sub 3}) takes place mainly as a result of ternary fissions in the nuclear reactors, of those which 75% are in gaseous form and 25% in liquid form. In the liquid phase, the tritium is transported to the pipes of the primary coolant toward condensed tanks or tanks of drainage excesses of radioactive equipment, located in external areas of a BWR, as well as to the processes of radioactive wastes to be able to be directed to the liquid effluents. For that reason, is necessary to know the possible routes of the transport and processes of the tritium in a BWR to control this radioisotope in the site of the event of leaks in equipment s and buried pipes, avoiding that emigrates toward underground flowing and an impact to the environment and to the people in general. (Author)

  5. Instrumenting a pressure suppression experiment for a MK I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    A scale test facility of a pressure suppression system from a boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy experimental data during a hypothetical loss-of-coolant accident. The instrumentation verified the analysis of the dynamic loading of the pressure suppression system

  6. Pulsation characteristics of boiling water cooled reactor two fuel assembly model

    International Nuclear Information System (INIS)

    The results of experimental studies into the pulsation characteristics of the natural circulation circuit model for the boiling water cooled reactor are given. Influence of nonidentity of fuel assembly power on stability of coolant flow rate was investigated. The methods for avoiding the whole circuit and interassembly hydrodynamic instabilities are suggested

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

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

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

  8. SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

    International Nuclear Information System (INIS)

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

  9. SWR 1000: A Next-Generation Boiling Water Reactor Ready for Deployment

    International Nuclear Information System (INIS)

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 design is particularly suitable for countries whose power systems do not include any large power plants. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (IandC) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10x10 rod array to a 12x12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free start-up, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving

  10. Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

    International Nuclear Information System (INIS)

    Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

  11. Observation of critical heat flux mechanism in horizontal pool boiling of saturated water

    International Nuclear Information System (INIS)

    Highlights: • A large dry patch is formed under a coalesced massive bubble. • A residual dry patch is produced at the departure of the coalesced massive bubble. • The residual dry patch expands to larger one due to enhanced nucleation activity. • Continuous re-expansion of the residual dry patch triggers the boiling crisis. - Abstract: We observed the global boiling structure and dynamic behavior of dry areas in a synchronized manner to identify the critical heat flux (CHF) triggering mechanism in a horizontal pool boiling of saturated water. A transparent Indium Tin Oxide (ITO) heating surface was used to accommodate a total reflection technique. The total reflection images captured the detailed processes of the generation of discrete dry spots, the formation and rewetting of large dry patches, and the irreversible expansion of the dry patch which led to the occurrence of the CHF. Contrary to the common postulation that a thin liquid film exists stably under a coalesced massive bubble, the base of hovering massive bubbles was almost dry at more than 10% below the CHF condition. The key element that determines the occurrence of the CHF was the production of the residual dry patch and consecutive re-expansion of the residual dry patch owing to the enhanced bubble nucleation activity with an increase in wall superheat, rather than the complete dryness of a boiling surface under a hovering massive bubble. The dry area fraction of the present water boiling test was similar to that of the R-113 boiling test in the literature in spite of significant differences in the wettability and physical properties

  12. The chemistry of feedwater for boiling-water and pressurized-water reactors

    International Nuclear Information System (INIS)

    In a nuclear power plant the purity of the feedwater depends largely on whether a condensate polishing plant is provided, whether the loop is conditioned and on the presence of corrosion products originating in the materials from which the loop is made. The feedwater specification depends on the type of steam generator used. The article defines the characteristic parameters of a condensate polishing plant (CPP), such as the 'degree of polishing' and 'practical exchange capacity of the resins' and indicates how they can be determined. In pressurized-water reactors (PWR) the feedwater is normally conditioned with hydrazine. Measurements are quoted to demonstrate that, in contrast to conventional plants, the point of injection is immaterial as regards the copper content of the feedwater. Moreover, the iron content of the feedwater of a PWR can be reduced by using cyclic amines. The feedwater chemistry of a BWR is discussed by referring to oxygen, iron and copper measurements. The authors show that in loops in which the feed-heater condensate is pumped forwards and where a feedwater tank is provided, the stipulated purity of the feedwater can be attained by suitable measures (such as mechanical filtration, prevention of erosion-corrosion, and so on). (Auth.)

  13. ROSA-III experimental program for BWR LOCA/ECCS integral simulation tests

    International Nuclear Information System (INIS)

    This is the final report of the ROSA-III experimental program, in which the summary of integral simulation test results is described on thermal-hydraulic behavior during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) and on the effectiveness of the emergency core cooling system (ECCS). Also presented in the report are the assessment results of computer codes for the BWR LOCA analysis and of the similarity between ROSA-III test results and thermal-hydraulic phenomena during a BWR LOCA by using ROSA-III test data and code analysis results. The ROSA-III facility is a volumetrically scaled (1/424) BWR system with an electrically heated core consisting of four half-length bundles. Many test series were conducted between April 1978 and March 1983. The similarity between a ROSA-III test and a BWR LOCA concerning the fundamental thermal-hydraulic phenomena has been confirmed for major ROSA-III tests. The accident scenario has been well understood and defined for various break locations and break sizes. The effectiveness of the current BWR ECCS design has been well demonstrated. (author)

  14. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use of two fuel concepts based on a mixed oxide (MOX) fuel and an innovative thorium-plutonium (ThPu) fuel is investigated by a developed simulation model encompassing thermal hydraulics, neutronics, and fuel burnup. The main feature of these fuel concepts is the axially varying enrichment in plutonium which is, in this work, recycled from spent nuclear fuel and shows a high fission fraction of the absorption cross section for fast incident neutron energies. The potential of balancing the overall fuel utilization by an increase of the fission rate in the upper part of the active height with a combination of the harder spectrum and the higher fission fraction of the absorption cross section in the BWR core is studied. The three particular calculational models for thermal hydraulics, neutronics, and fuel burnup provide results at fuel assembly and/or at core level. In the former case, the main focus lies on the thermal hydraulics analysis, fuel burnup, and activity evolution after unloading from the core and, in the latter case, special attention is paid to reactivity safety coefficients (feedback effects) and the optimization of the operational behavior. At both levels (assembly and core), the isotopic buildup and depletion rates as a function of the active height are analyzed. In addition, a comparison between the use of conventional fuel types with homogeneous enrichments and the use of the innovative fuel types is made. In the framework of the simulations, the ThPu and the MOX

  15. Determination of critical parameters during boiling water flow through cylindrical channels

    International Nuclear Information System (INIS)

    The critical flow of subcooled water through long 10≤L/D<20 and short L/D=0.5 and L/D=2.5 cylindrical cannels (packings) with sharp inlet edge is studied. The flow rate, critical pressure and reactive forces are determined. The nomogram for determination of critical pressure ratioes depending on initial pressure and initial temperature of boiling water and the formulae for this dependence calculation are built

  16. Flow processes during subcooled boiling in fuel rod clusters of water-cooled reactors

    International Nuclear Information System (INIS)

    The theoretical fundamentals for the thermohydraulic calculation of fuel rod clusters in light water-cooled reactors are presented with special regard to boiling on fuel rods in unsaturated water. It is shown which preconditions concerning the structure of the two-phase flow must be met in order to apply the methods of single-phase continuum mechanics to two-phase flows. (orig./TK)

  17. Development of evaluation tool for radiation dose rate distribution in PCV of Hamaoka BWR plants based on water chemistry

    International Nuclear Information System (INIS)

    We have developed an evaluation tool for the radiation dose rate distribution of the work areas in the primary containment vessel (PCV) of Units 3, 4 (BWR5) and 5 (ABWR) at Hamaoka NPS. This tool has been constructed based on the transport behavior of radioactive corrosion products in the primary cooling water of BWR. This tool can be used to evaluate quantitatively the effects of the dose reduction methods by water chemistry control or radiation management. It is composed of two calculation codes; water chemistry code (ACTTUBE) and radiation dose rate code (RADTUBE). ACTTUBE calculates the piping dose rates based on the mass balance of corrosion products, 6 kinds of metal and 5 kinds of radionuclide, among the parts of primary cooling water, such as reactor water, feed water, fuel rod surface and out-of-core piping surface. RADTUBE calculates the dose rate distribution based on the radiation shielding calculation from a calculation result of ACTTUBE. Additionally, this tool has a visualization function of calculated radiation dose rate distribution in the PCV by using a wireless controller and 3D glasses/monitor in order to improve user convenience. The accuracy of the tool's calculation results was evaluated using the water chemistry data and radiation dose rate data of the Hamaoka plants. As a result, it was confirmed that this tool had sufficient accuracy to be used in the evaluation of radiation dose rates for the radiation management of actual plants. (author)

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

    International Nuclear Information System (INIS)

    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 B4C-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

  19. Bubble Behavior in Subcooled Pool Boiling of Water under Reduced Gravity

    Science.gov (United States)

    Suzuki, Koichi; Suzuki, Motohiro; Takahash, Saika; Kawamura, Hirosi; Abe, Yoshiyuki

    2003-01-01

    Subcooled pool boiling of water was conducted in reduced gravity performed by a parabolic flight of aircraft and a drop-shaft facility. A small stainless steel plate was physically burned out in the subcooled water by AC electric power during the parabolic flight. Boiling bubbles grew with increasing heating power but did not detached from the heating surface. The burnout heat fluxes obtained were 200 ~ 400 percent higher than the existing theories. In the ground experiment, boiling bubbles were attached to the heating surface with a flat plate placed over the heating surface, and the experiment was performed by the same heating procedure as practiced under the reduced gravity. Same burnout heat fluxes as under the reduced gravity were obtained by adjusting the plate clearance to the heating surface. As the heating time extended longer than the reduced gravity duration, the burnout heat fluxes decreased gradually and became constant. Contact area of bubbles with heating surface was observed using a transparent heating surface in microgravity performed by a drop-shaft facility. The contact area of bubbles increased significantly at the start of microgravity. It is suggested by the experimental results that the boiling bubbles expand rapidly in the high heat flux region and the rapid evaporation of liquid layer remained between the bubbles and the heating surface raises up the critical heat flux higher than the existing theories in microgravity.

  20. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R. [AREVA, AREVA NP, Erlangen (Germany)

    2008-07-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  1. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  2. Importance of momentum effects in BWR reactor vessel modeling

    International Nuclear Information System (INIS)

    The most severe class of events (other than design-basis LOCA's) for boiling water reactors (BWR's) entails rapid pressurization of the reactor vessel (RV) due to rapid valve closures in the main steamlines. The severity of these events is caused by the fact that the pressurization of the core results in a rapid decrease in the core void fraction and, coupled to a strong negative void coefficient of reactivity, in a rapid power increase if the reactor protective system is slow in responding. The DYNODE-B program models the nuclear steam supply system of BWR's. Early versions only explicitly calculated the RV dome pressure. The latest version permits optional explicit calculation of the dome, core outlet, and core average pressures. The use of the new option has shown the influence of the two-phase momentum effects on the core pressure and flow transients resulting from pressure wave reflections at the liquid (incompressible) core inlet region

  3. Diagnosis of nonlinear BWR oscillations using TRAC/BF1

    International Nuclear Information System (INIS)

    The nonlinear nature of boiling water reactor (BWR) stability has been demonstrated in both experimental tests and lumped parameter calculational models. Point kinetic reactivity feedback is nonlinear because of its functional dependence on fuel temperature and moderator density. The TRAC/BF1 model used in this analysis differs from a lumped parameter model in its spatial extent. The model, intended to be consistent with a BWR/4, was developed with four active fuel channel components representing one hot, two average, and one peripheral bundles. The vessel internals were modeled explicitly. These internals include lower and upper plena, separator/dryers, core shroud, and dryer skirt. The jet pump/recirculation system is modeled in an azimuthally symmetric fashion. The feedwater and steam line boundary conditions are based on time-dependent data representative of that observed during the LaSalle oscillation event

  4. Study of a Heavy-Water Reactor with Boiling Heavy-Water Coolant

    International Nuclear Information System (INIS)

    Among the possible types of heavy-water reactor, those cooled by heavy water would appear to combine the advantages of excellent neutron economy and a well-tried cladding material; this allows optimum utilization of uranium under the present conditions of technology. Placing the reactor, the handling equipment, and the heat exchangers together in a prestressed concrete vessel appreciably simplifies operating problems by reducing the number of hermetic seals in contact with the pressurized heavy water. This arrangement is only effective if a large proportion of the heat transfer is by phase change, so as to keep the amount of coolant to a minimum. The Commissariat à Energie Atomique has made a study of a boiling heavy-water reactor under a co-operation agreement with the Siemens and Sulzer Companies and with the participation of the Socia Company. The paper describes the main features of these projects as well as the main technological problems raised by this design which relate to the thermal insulation of the concrete vessel in the presence of a two-phase fluid; the handling equipment which must function in steam at 300°C; and the accessibility of the exchangers. (author)

  5. Co-boiling of NAPLs and water during thermal remediation: experimental and modeling study

    Science.gov (United States)

    Krol, M.; Zhao, C.; Mumford, K. G.; Sleep, B. E.; Kueper, B. H.

    2015-12-01

    The persistence of non-aqueous-phase liquids (NAPLs) in the subsurface has led to the development of several remediation technologies to address this environmental problem. One such group of technologies (in situ thermal treatment) uses heat to volatilize contaminants. Subsurface temperature measurements are often used to monitor progress and optimize contaminant removal. However, when NAPL and water are heated together, gas is created at a temperature lower than the boiling point of either liquid (co-boiling), which can affect temperature observations. To examine the effect of co-boiling on observed temperatures and NAPL mass removal, a series of heated laboratory experiments were performed using single and multi-component NAPLs. The experiments consisted of glass jars filled with a mixture of sand, water, and NAPL mixed to obtain an approximately uniform NAPL distribution within the jar. The experiments were heated from the outside and interior temperatures were measured using a thermocouple. The tests showed that local-scale temperature measurements are unreliable in indicating the end of co-boiling and may not indicate complete mass removal. This is because a well-defined co-boiling plateau does not exist when heating a multi-component NAPL and the temperature is dependent on the proximity of NAPL to the monitoring point. To further investigate temperature distributions and the potential to use gas production as a complementary indicator of NAPL removal, a 2D finite-difference mass transport model was used that incorporated heat transport, latent heat, phase change, and a multicomponent gas phase and used a macroscopic invasion percolation (MIP) model to simulate gas movement. Latent heat was calculated by multiplying specific latent heat, which is an intrinsic property of a substance, by the amount of liquid mass being vaporized and its incorporation into the model allowed for the simulation of co-boiling plateaus (during single component NAPL boiling). The

  6. 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.

  7. Reliability of BWR high pressure core cooling

    International Nuclear Information System (INIS)

    The high pressure coolant injection system (HPCI), and the reactor core isolation cooling system (RCIC) are steam turbine driven systems that can inject water into a boiling water reactor at full operating pressure. Their purpose is to supply water during any failure that allows water to be lost while the reactor is at pressure and temperature. A large number of BWR plants are not meeting HPCI and RCIC performance goals for core cooling. NSAC considers concurrent failure of NPCI and RCIC to be the most probable potential cause of low reactor water level and possibly fuel damage in a boiling water reactor. Between January 1978 and May 1981, 169 licensee event reports were filed where HPCI or RCIC was inoperable or was declared inoperable. The present effort has shown that at least 40% of NPCI and RCIC problems might be averted by a high quality preventive maintenance program. About half of the plants do not perform cold quick-start surveillance testing of HPCI and RCIC. They do perform routine startup tests, but the equipment is first preheated and the startup is relatively gentle. However, emergency start-ups are abrupt and from the cold condition. Therefore, cold quick-start testing is the only way to assure that all components, control systems, and instruments are functioning correctly for automatic safety initiation. (author)

  8. Development of large-capacity main steam isolation valves and safety relief valves for next-generation BWR plant

    International Nuclear Information System (INIS)

    A study was made of high capacity main steam isolation valves (MSIV) and safety relief valves (SRV) for the main steam line of a boiling water reactor (BWR). The next-generation BWR plants, which are planned to have higher thermal power, have raised concerns relating to the main steam line of an increase in maintenance work to SRVs and erosion of the MSIV valve seat due to the increased main steam flow velocity. In this research project, the capacity of the MSIV and SRV was increased and the valve configuration was changed in an attempt to solve these problems. (authors)

  9. Boiling-up of liquid nitrogen jet in water

    Science.gov (United States)

    Nakoryakov, V. E.; Tsoi, A. N.; Mezentsev, I. V.; Meleshkin, A. V.

    2014-06-01

    The hydrodynamic processes occurring at injection of cryogenic liquid into water pool were studied experimentally. Processes accompanying the phase transitions were registered. Data testify the developing pressure burst with an amplitude sufficient for possible formation of gas hydrates when methane is injected as a cryogenic fluid.

  10. Dynamic safety systems in BWR plant safety systems

    International Nuclear Information System (INIS)

    Dynamic Safety Systems (DSSs) are reactor safety function systems that are functionally controlled using dynamic rather than static processes. All components including software, whose failure could result in a critical safety system failure, are operationally verified by hard-wired components. Dynamic Safety Systems have been enveloped in the United Kingdom by AEA Technology for use in gas cooled reactors. One such system, known as ISAT trademark, is described in this paper. Through use of scenario testing of a DDS emulator on a Boiling Water Reactor plant training simulator described in this paper. Through use of scenario testing of a DSS emulator on a Boiling Water Reactor plant training simulator, it is shown that a DSS can provide a cost effective safety system in BWR power plants

  11. CHF Enhancement of SiC-water nanofluids in Pool Boiling Experiment

    International Nuclear Information System (INIS)

    SiC nanofluids were used for Critical heat flux(CHF) enhancement in the case of water pool boiling. Many kinds of nanofluids have been highlighted as a simple way to gain high thermal performance of fluids. Also, one of the ceramic particle, SiC is received attention these days as a promising material because of its relatively high thermal properties. In this study, SiC nanofluids were investigated to measure its thermal performance in water pool boiling experiment especially for CHF. The volume concentration of SiC nanofluids were 0.0001%, 0.001%, 0.01%. Several characteristic of SiC nanofluids, such as Zeta potential, and contact angle which could be affect on thermal performance of the fluids had been measured. The experiments were conducted under atmospheric pressure. The CHF has been enhanced upto 53.1% at volume concentration 0.01% SiC nanofluids

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

    International Nuclear Information System (INIS)

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

  13. Ion chromatography campaigns at several boiling water reactors

    International Nuclear Information System (INIS)

    Water chemistry characterization campaigns using on-line ion chromatography were conducted at several BWRs during 1989. Some of the highlights from these campaigns, along with equipment and IC methods are presented in this paper. Monitoring copper ion in final feedwater and filter demineralizer effluents at levels <0.2 ppb enabled optimum operation of the condensate demineralizer system. The need for new precoats was based on ion trends, not pressure drop or volume limitations. Conductivity transients, due to power reductions, were characterized. Magnesium, calcium, sulfate and chromate were the major contributors. Start up, water chemistry transients, were monitored, revealing organics from sealing compounds used in the condenser and turbine. On-line IC is an effective tool for understanding power plant chemistry and improving operations of condensate cleanup systems

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

    OpenAIRE

    Fridström, Richard

    2010-01-01

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

  15. Simulation and fault-detection of a pressure control servosystem in a Boiling Water Reactor

    International Nuclear Information System (INIS)

    This master thesis describes a Simnon model of a boiling water reactor to be used in simulating faults and disturbances. These faults and disturbanses will be detected by noise analysis. Some methods in identification and noise analysis are also described and are applied on some malfunctions of a servo. A Pascal program for recursive parameter identification was also written and tested. This program is to be used in an expert system for noise analysis on the nuclear power plant Barsebaeck. (author)

  16. BOILING OF WATER AND ORGANIC LIQUIDS ON LOW-TEMPERATURE POROUS SURFACES OF HEAT PIPES

    OpenAIRE

    Шаповал, Андрій Андрійович; Панов, Євген Миколайович; Сауліна, Юлія Валеріївна; Романчук, Борис Васильович; Трубійчук, Р. П.

    2015-01-01

    The experimental study results of the influence of porous metal fiber structures on the intensity of two-phase heat transfer of water and acetone boiling on porous surfaces in conditions of free movement and capillary transport of liquids are presented in the article. The experiments were realized using specially designed experimental installation simulated the operating conditions of heat pipes and thermosyphons. Such conditions are typical for two-phase heat transfer devices – heat pipes an...

  17. Evaluation of damages of airplane crash in European Advanced Boiling Water Reactor (EU-ABWR)

    International Nuclear Information System (INIS)

    European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash (APC), severe accident mitigation systems, N+2 principle in safety systems and a large output of 1600 MWe. Thanks to above mentioned features, EU-ABWR's design objectives and principles are consistent with safety requirements in an European market. In this paper, evaluation of damages induced by APC has been summarized. (author)

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

    OpenAIRE

    Hsoung-Wei Chou; Chin-Cheng Huang

    2015-01-01

    The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to ...

  19. Pool boiling heat transfer of water/ γ-alumina micro-fluids around the horizontal cylinder

    Science.gov (United States)

    Nikkhah, V.; Hormozi, F.

    2016-04-01

    A set of experiments was performed to quantify the pool boiling heat transfer coefficient of water/ γ-alumina micro-fluids at mass concentration ranged from 0.1 to 0.4 % of micro-particles with mean size of 1-2 μm. To stabilize the prepared micro-fluid, pH control, stirring and adding the SDS as a surfactant were carried out. Also, thermal conductivity of micro-fluids are measured using KD2 decagon pro. Results showed that micro-fluids have relatively higher thermal conductivity rather than the base fluids. According to the results, there are two distinguishable heat transfer regions namely natural convection and nucleate boiling regions. Influence of some operating parameters such as heat flux, mass concentration of micro-particles and surface fouling resistance on the pool boiling heat transfer coefficient were experimentally studied and briefly discussed. Results demonstrated a significant deterioration of heat transfer coefficient of micro-fluids in comparison with the base fluid over the extended time (1000 min of operation) in nucleate boiling region, while in natural convection region, enhancement of heat transfer coefficient is registered. According to the results, heat transfer coefficient is strongly controlled by/ γ-alumina concentration due to the deposition of micro-particles on the heating section. Rectilinear changes of scale formation with time in term of fouling resistance were clearly seen at regions, where natural convection is a dominant heat transfer mechanism and also for higher heat fluxes at nucleate boiling heat transfer region.

  20. Safety implications of an integrated boiling curve model for water-cooled divertor channels

    International Nuclear Information System (INIS)

    The international fusion community is actively researching advanced heat transfer methods for removal of high thermal loads from next-generation divertor assemblies. Such advanced techniques may indeed optimize the operational and economical performance of future divertor designs. However, with its extensive operational database, water-cooling remains as one of the optimum choices for near-term divertor designs. Critical heat flux (CHF) is the maximum heat flux that water, at a given set of inlet conditions, can remove via fully developed nucleate boiling. Accordingly, an accurate CHF calculation is of the utmost importance for maintaining adequate safety margins in divertor operation. This paper uses the integrated boiling curve model developed at Sandia National Laboratories to examine the safety implications of calculating the CHF. In particular, this paper focuses on the influence of the finite element peaking factor (FEPF) that converts the heat flux predicted by CHF correlations into a plasma heat flux that can be measured. The analyses illustrate that the FEPF is proportional to the plasma heat flux and thus accurate calculation of the CHF requires the use of the appropriate FEPF for the given water conditions and plasma heat flux. It is shown that using a geometric peaking factor is inadequate since the true peak factor is dependent upon the plasma heat flux. The conclusion is that a finite element analysis incorporating an integrated boiling curve is required for accurate calculation of the CHF

  1. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    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)

  2. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

    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.

  3. Natural convection type BWR reactor

    International Nuclear Information System (INIS)

    In a natural convection type BWR reactor, a mixed stream of steams and water undergo a great flow resistance. In particular, pressure loss upon passing from an upper plenum to a stand pipe and pressure loss upon passing through rotational blades are great. Then, a steam dryer comprising laminated dome-like perforated plates and a drain pipe for flowing down separated water to a downcomer are disposed above a riser. The coolants heated in the reactor core are boiled, uprise in the riser as a gas-liquid two phase flow containing voids, release steams containing droplets from the surface of the gas-liquid two phase, flow into the steam dryer comprising the perforated plates and are separated into a gas and a liquid. The dried steams flow to a turbine passing through a main steam pipe and the condensated droplets flow down through the drain pipe and the downcomer to the lower portion of the reactor core. In this way, the conventional gas-liquid separator can be saved without lowering the quality of steam drying to reduce the pressure loss and to improve the operation performance. (N.H.)

  4. Simulation of heat and mass transfer in boiling water with the Melodif code

    International Nuclear Information System (INIS)

    The Melodif code is developed at Electricite de France, Research and Development Division. It is an eulerian two dimensional code for the simulation of turbulent two phase flows (a three dimensional code derived from Melodif, ASTRID, is currently being prepared). Melodif is based on the two fluid model, solving the equations of conservation for mass, momentum and energy, for both phases. In such a two fluid model, the description of interfacial transfers between phases is a crucial issue. The model used applies to a dominant continuous phase, and a dispersed phase. A good description of interfacial momentum transfer exists in the standard MELODIF code: the drag force, the apparent mass force... are taken into account. An important factor for interfacial transfers is the interfacial area per volume unit. With the assumption of spherical gas bubbles, an equation has been written for this variable. In the present wok, a model has been tested for interfacial heat and mass transfer in the case of boiling water: it is assumed that mass transfer is controlled by heat transfer through the latent massic energy taken in the phase that vaporizes (or condenses). This heat and mass transfer model has been tested in various configurations: - a cylinder with water flowing inside, is being heated. Boiling takes place near the wall, while bubbles migrating to the core of the flow recondense. This roughly simulates a sub-cooled boiling phenomenon. - a box containing liquid water is depressurized. Boiling takes place in the whole volume of the fluid. The Melodif code can simulate this configuration due to the implicitation of the relation between interphase mass transfer and the pressure variable

  5. Impact of Plant Noise on BWR Stability Analyses

    International Nuclear Information System (INIS)

    A small amount of thermal-hydraulic noise is present in a boiling water reactor (BWR). The noise originates primarily from minor fluctuations in fluid flow and pressure distribution in the recirculation system of the BWR and manifests itself as a small fluctuation on the order of 1-2% for the average power range monitors (APRM) during normal operation. A larger noise level is observed for single-loop operation than for two-loop operation. This noise has an impact on the stability performance of the BWR. This is particularly the case when the noise contains a significant component at the resonant frequency for BWR instabilities, which is typically on the order of 0.5 Hz. For a pump trip event that can lead to instability, the noise will impact the growth rate of the reactor instability. The initial magnitude of the oscillations will be larger as the decay ratio increases above unity. For operation at low flow, such as for minimum pump speed or single loop operation where the decay ratio is larger, the impact of noise could lead to small oscillations at the resonant frequency for the APRM signals. The impact of noise on BWR instabilities is analyzed with the TRACG code. TRACG consists of a multi dimensional two-fluid thermal hydraulics model and the three-dimensional kinetics model consistent with the GE 3D core simulator, PANACEA. TRACG models the reactor primary system and has been extensively qualified against test data and BWR plant data. Thermal hydraulic instability test data, as well as data from BWR instability events and tests, have been used extensively in this qualification. This paper demonstrates the impact of noise on BWR stability response for events leading to instability, such as pump trip events, as well as operation at low core flow due to single loop operation. The impact is illustrated through sensitivity studies with the TRACG code and by comparison to plant data. The impact of reactor noise on the performance of the instability detection system

  6. Liquid-cooled nuclear reactor, especially a boiling water reactor

    International Nuclear Information System (INIS)

    A nuclear reactor with a special arrangement of fuel rods in the core is designed. Each fuel element has its shaft which is made of sheets, has the same cross section as the fuel element and protrudes at least the length of the control rod above the reactor core. Made of a zirconium alloy in the core area and of stainless steel above it, the shaft is equipped with channels for sliding the rods in and out and serves to spatially secure the position of the rods. Coolant flow is provided by the chimney effect. The shaft can conveniently enclose the control rod drive. It can also serve to bear the water separator. Moreover, it can constitute a part of the casing which surrounds the fuel rods and keeps the fuel in an intimate contact with the coolant; the other part of this casing is constituted by inserted sheets which can conveniently have the shape of angles. The walls of neighboring shafts form a compartment accommodating a neutron absorber plate. (M.D.). 11 figs

  7. Increased fuel column height for boiling water reactor fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Matzner, B.

    1993-06-15

    Rods to maintain said fuel rods upstanding and permitting the exit of water and generated steam is described; a fuel bundle channel extending from said lower tie plate to the vicinity of said upper tie plate, and surrounding said fuel rods therebetween for producing an isolated flow region through said matrix of upstanding fuel rods for the generation of steam by nuclear reaction within said fuel rods; a first plurality of said fuel rods being full length fuel rods for extending fully between said upper and lower tie plates; and, a second plurality of said fuel rods being part length fuel rods for extending part way from a supported disposition on said lower tie plate to a point of fuel rod termination below said upper tie plate whereby a vacated vertical interval is defined between the upper end of said part length fuel rod and said upper tie plate; the improvement to said first plurality of full length fuel rods comprising in combination: said full length fuel rods including a first lower region having a first and smaller diameter containing said pellets of fissionable material; and, at least some of said full length fuel rods including an upper region containing said plenum which is devoid of fuel pellets having a second and larger diameter for providing to said plenum an expanded volume whereby the flow area overlying said part length fuel rods defines additional outflow area adjacent said plenums and the active length of fissionable pellets within said full length fuel rods can be increased.

  8. Increased fuel column height for boiling water reactor fuel rods

    International Nuclear Information System (INIS)

    Rods to maintain said fuel rods upstanding and permitting the exit of water and generated steam is described; a fuel bundle channel extending from said lower tie plate to the vicinity of said upper tie plate, and surrounding said fuel rods therebetween for producing an isolated flow region through said matrix of upstanding fuel rods for the generation of steam by nuclear reaction within said fuel rods; a first plurality of said fuel rods being full length fuel rods for extending fully between said upper and lower tie plates; and, a second plurality of said fuel rods being part length fuel rods for extending part way from a supported disposition on said lower tie plate to a point of fuel rod termination below said upper tie plate whereby a vacated vertical interval is defined between the upper end of said part length fuel rod and said upper tie plate; the improvement to said first plurality of full length fuel rods comprising in combination: said full length fuel rods including a first lower region having a first and smaller diameter containing said pellets of fissionable material; and, at least some of said full length fuel rods including an upper region containing said plenum which is devoid of fuel pellets having a second and larger diameter for providing to said plenum an expanded volume whereby the flow area overlying said part length fuel rods defines additional outflow area adjacent said plenums and the active length of fissionable pellets within said full length fuel rods can be increased

  9. Contingency programs for BWR pipe cracking

    International Nuclear Information System (INIS)

    General Electric (GE) has aggressively addressed the problem of Intergranular Stress Corrosion Cracking (IGSCC). Intergranular Stress Corrosion Cracking has occurred in boiling water reactors in less than 1% of the Type-304 stainless steel welds in operating plants. However a comprehensive program by GE, with EPRI support in many cases, has provided technical solutions to limit IGSCC. As part of this program the Nuclear Services Department (NSD) has taken new technology and applied it to programs for the prevention of IGSCC and for repair of pipe cracks should they occur. The purpose of this paper is to provide a description and the current status of the major ongoing service programs for dealing with the problems related to IGSCC in BWR operating plants

  10. Potential uses of high gradient magnetic filtration for high-temperature water purification in boiling water reactors

    International Nuclear Information System (INIS)

    Studies of various high-temperature filter devices indicate a potentially positive impact for high gradient magnetic filtration on boiling water reactor radiation level reduction. Test results on in-plant water composition and impurity crystallography are presented for several typical boiling water reactors (BWRs) on plant streams where high-temperature filtration may be particularly beneficial. An experimental model on the removal of red iron oxide (hematite) from simulated reactor water with a high gradient magnetic filter is presented, as well as the scale-up parameters used to predict the filtration efficiency on various high temperature, in-plant streams. Numerical examples are given to illustrate the crud removal potential of high gradient magnetic filters installed at alternative stream locations under typical, steady-state, plant operating conditions

  11. State of the art report on boiling water reactor stability (SOAR on BWRs)

    International Nuclear Information System (INIS)

    Starting issues of this SOAR are BWR plant descriptions including peculiarities relevant to stability and the manifestation of instabilities during operation. The report continues with the characterization of instabilities from various experiments, the features and the capabilities of relevant codes and models, BWR core instrumentation and control, the stability behaviour of operating BWR plants and the regulatory approach to the stability issue. The main conclusion is that the BWR stability should not be considered as a safety issue; however R and D in specific areas is recommended

  12. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

    International Nuclear Information System (INIS)

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

  13. On recriticality during reflooding of a degraded boiling water reactor core

    International Nuclear Information System (INIS)

    In-vessel core melt progression in Nordic BWRs has been studied as a part of the RAK-2 project within the Nordic Nuclear Safety Programme 1994-1997. A part of this study was the evaluation of possibility and consequences of recriticality in a re-flooded, degraded BWR core. The objective of the study was to examine, if a BWR core in a Nordic nuclear power plant can reach critical state in a severe accident, when the core is re-flooded with un-borated water from the emergency core cooling system and what is the possible power augmentation related to recriticality. The containment response to elevated power level and consequent enhanced steam production was evaluated. The first sub-task was to upgrade the existing neutronics/thermal hydraulic models to a level needed for a study of recriticality. Three different codes were applied for the task: RECRIT, SIMULATE-3K and APROS. Preliminary calculations were performed with the three codes. The results of present studies showed that reflodding of a partly control rod free core gives a recriticality power peak of a substantial amplitude, but with a short duration due to the Doppler feedback. The energy addition is small and contributes very little to heat-up of the fuel. However, with continued reflodding the fission power increases again and tend to stabilize on a level that can be ten per cent or more of the nominal power, the level being higher with higher reflooding flow rate. A scoping study on TVO BWR containment response to a presumed recriticality accident with a long-term power level being 20% of the nominal power was performed. The results indicated that containment venting system would not be sufficient to prevent containment overpressurization and containment failure would occur about 3-4 h after start of core reflooding. In the case of station blackout with operating ADS the present boron system would be sufficient to terminate the criticality even prior to containment failure, but in case of feedwater LOCA and

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

    International Nuclear Information System (INIS)

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

  15. Cyclic crack growth through typical weld HAZ microstructures of SA 533 gr.B steel in simulated BWR water environment

    International Nuclear Information System (INIS)

    The effect of BWR water environment on the cyclic crack growth characteristics of SA 533 gr.B steel specimen of a common heat with different microstructures was examined with major emphasis placed on the crack growth along the weld heat affected zone (HAZ). The variation employed was provided by three different heat treatments, i.e. (1) air-quenched and tempered, (2) oil-quenched and tempered and (3) as oil-quenched, each of which were to simulate the localized variability of typical alternated microstructures of the weld-heat affected zones appearing along fusion lines. In the preliminary air tests, no essential difference was observed among all kinds of materials in their δK versus da/dN relationships for two stress ratio conditions, i.e. R = 0.1 and 0.5. In the BWR water environment, the crack growth rates of those three types of materials were accelerated relative to the air environment results. The degrees of acceleration were not greatly different to each other under stress ratio, R = 0.1, while considerable difference was noted in case of higher stress ratio, R = 0.5. Under the high stress ratio test condition, the quenched material, which was with martensitic structure and hardening, showed highest degree of environment sensitivity among others. The obtained results were consistent with the cyclic crack growth rates when cracks were advancing along weld fusion lines through HAZ, where considerable level of residual stresses were to have remained. (author)

  16. Method for increasing the stability of a boiling water cooled reactor with natural coolant circulation and a boiling water cooled reactor with natural coolant circulation (its versions)

    International Nuclear Information System (INIS)

    The invention is aimed at improving the safety of a boiling water reactor with natural coolant circulation and increasing the reactor core power density by increasing the coolant flowrate and neutron flux stability as well as by reducing the medium compressibility effectiveness in pressure compensator in dynamic modes. The reactor vessel includes the core, draught section, heat exchangers and a pressure compensator. A part of the pressure compensator is separated by a barrier with calibrated openings possessing a limited capacity and hydrolocks. The calibrated openings in the barrier are located below the coolant level and a part of space separated by a barrier is filled with gas from external system. The part of the barrier projecting above the coolant level is adjacent to heat exchangers. In transitional regimes with the change of pressure in the circulation circuit a hydrolock facilitates to reactor vessel projection against repressing and keeps the barrier from excessive power load

  17. The Behavior of Corrosion Products in Sampling Systems under Boiling Water Reactor Conditions

    International Nuclear Information System (INIS)

    A high pressure loop has been used to simulate sampling systems employed under BWR conditions. The reliability of the sampling method was studied in a series of six test runs. A variety of parameters that are thought to influence the reliability of the sampling was investigated. These included piping geometry, water oxygen content, flow, temperature and temperature gradients. Amongst other things the results indicate that the loss by deposition of iron containing corrosion products does not exceed 50 %; this figure is only influenced to a minor extent by the above mentioned parameters. The major part of the corrosion products thus deposited is found along the first few meters of the piping and cooler coil. A moderate prolongation of a pipe which is already relatively long should thus be incapable of producing a major influence on the sampling error

  18. Severe accident mitigation features of the economic simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper provides an overview of the Economic Simplified Boiling Water Reactor (ESBWR)severe accident mitigation systems. The major severe accident types are described and the systems credited for mitigating the severe accidents are discussed, including the Basemat Internal Melt Arrest Coolability (BiMAC) device, the Passive Containment Cooling System (PCCS), and the advantages of suppression pool water for scrubbing during containment venting. The ruggedness of the containment and reactor building designs for accommodating beyond design accident conditions is also discussed. (author)

  19. A potential of boiling water power reactors with a natural circulation of a coolant

    International Nuclear Information System (INIS)

    The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

  20. Water boiling on the corium melt surface under VVER severe accident conditions

    International Nuclear Information System (INIS)

    Experimental results are presented on the interaction between corium melt and water supplied onto its surface. The tests were conducted on the Rasplav-2' experimental facility. Induction melting in a cold crucible was used to produce the melt. The following data have been obtained: heat transfer at water boiling on the melt surface, aerosol release, structure of the post-interaction solidified corium. The corium melt had the following composition, mass %: 60%UO2- 16%ZrO2- 15%Fe2O3 - 6%Cr2O3-3%Ni2O3. The melt surface temperature was 1650-1700degC. (author)

  1. Enhancement of flow boiling of subcooled water on transverse ribbed surface

    International Nuclear Information System (INIS)

    The present work deals with enhancement of flow boiling heat transfer using transverse ribs of rectangular cross section attached on a flat heating surface. High flux cooling is envisioned in the field of the leading edge technology such as high power laser applications, advanced metallurgical processes, fusion reactors, integrated circuit chips and so forth. In such advanced cooling devices subcooled boiling of water at high velocity is expected as one of most efficient and convenient means for heat removal exceeding 107W/m2. In the present experiment, a sheet of stainless-steel (10mm wide, 0.2mm thick and 80mm in heated length) was flush mounted on one wall of a vertical rectangular channel (a cross-section 20mm x 30 mm) and used as a heating surface by passing a direct current. Transverse-ribs made of plastic plate (0.5mm thick and 30mm in lateral length) were attached at an equal longitudinal spacing on the heating surface. Longitudinal spacing of ribs was varied from 2.5, 5, 10, 20mm to infinity (a flat surface without ribs), and the rib height was 2.5mm and 5.0mm. Experiments were conducted with water at a pressure of 0.12MPa in the range of mass velocity from 500kg/m2s to 2,000kg/m2s (water velocity from 0.5m/s to 2m/s) and subcooling from 20K to 50K. It was found that the ribbed structure strongly affects heat transfer both in the non-boiling and partial nucleate boiling regimes. Enhancement rate of heat transfer coefficient varied from 5% to 50% in excess compared with that for the flat heating plate. According to visual observation of boiling bubbles the circulating flow occurred in a space between each consecutive ribs and seemed to enhance heat transfer. For narrow rib spacing a large coalesced bubble filled up each rib space and impeded the exchange of vapor with liquid, leading to heat transfer deterioration. In the present experimental range of the ribs, a 50% increase of heat transfer coefficient was attained on the ribbed surface with 5mm height

  2. Pressure drop, heat transfer, critical heat flux, and flow stability of two-phase flow boiling of water and ethylene glycol/water mixtures - final report for project "Efficent cooling in engines with nucleate boiling."

    Energy Technology Data Exchange (ETDEWEB)

    Yu, W.; France, D. M.; Routbort, J. L. (Energy Systems)

    2011-01-19

    Because of its order-of-magnitude higher heat transfer rates, there is interest in using controllable two-phase nucleate boiling instead of conventional single-phase forced convection in vehicular cooling systems to remove ever increasing heat loads and to eliminate potential hot spots in engines. However, the fundamental understanding of flow boiling mechanisms of a 50/50 ethylene glycol/water mixture under engineering application conditions is still limited. In addition, it is impractical to precisely maintain the volume concentration ratio of the ethylene glycol/water mixture coolant at 50/50. Therefore, any investigation into engine coolant characteristics should include a range of volume concentration ratios around the nominal 50/50 mark. In this study, the forced convective boiling heat transfer of distilled water and ethylene glycol/water mixtures with volume concentration ratios of 40/60, 50/50, and 60/40 in a 2.98-mm-inner-diameter circular tube has been investigated in both the horizontal flow and the vertical flow. The two-phase pressure drop, the forced convective boiling heat transfer coefficient, and the critical heat flux of the test fluids were determined experimentally over a range of the mass flux, the vapor mass quality, and the inlet subcooling through a new boiling data reduction procedure that allowed the analytical calculation of the fluid boiling temperatures along the experimental test section by applying the ideal mixture assumption and the equilibrium assumption along with Raoult's law. Based on the experimental data, predictive methods for the two-phase pressure drop, the forced convective boiling heat transfer coefficient, and the critical heat flux under engine application conditions were developed. The results summarized in this final project report provide the necessary information for designing and implementing nucleate-boiling vehicular cooling systems.

  3. Space–time convergence analysis on BWR stability using TRACE/PARCS

    International Nuclear Information System (INIS)

    Highlights: ► Quantify TRACE/PARCS space–time discretization error for BWR stability prediction. ► Establish space and time discretization necessary for space–time converged models. ► Show that the space–time converged model gives more reliable results for both stable and unstable reactor. ► Use of the space–time converged model increases confidence in the prediction of BWR stability. -- Abstract: Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.

  4. Observation of contact area of bubbles with heating surface in pool boiling of water under microgravity

    International Nuclear Information System (INIS)

    Burnout heat flux was measured in subcooled pool boiling of water under attached boiling bubbles on heating surface with bubble holding plate in ground experiment. A thin stainless flat plate was employed for heating surface. The experimental setup and the heating procedures were same as used in reduced gravity experiment performed by a parabolic flight of jet aircraft. Same burnout heat flux as in the reduced gravity was obtained by adjusting the clearance between the bubble holder and the heating surface. They were 100 ∝ 400 percent higher than the widely accepted existing theories. As extending heating time longer than the reduced gravity duration until burnout occurred, burnout heat flux decreased gradually and became a constant value calculated from the existing theories. In a result of observing contact area of boiling bubbles with transparent heating surface, the contact area was smaller in quick heating time than that in long time heating at same heat flux. The experimental results suggest in microgravity that liquid layer is remained between rapidly expanded bubbles and heating surface. In microgravity experiment by a drop shaft facility, contact area of bubbles with heating surface increased considerably at starting of microgravity. (orig.)

  5. Analysis of mixed oxide fuel behavior under reduced moderation boiling water reactor conditions with FRAPCON-EP

    International Nuclear Information System (INIS)

    FRAPCON-EP models have been extended to better represent mixed oxide steady state fuel behavior under the Reduced moderation Boiling Water Reactor (RBWR) conditions. RBWR fuel is designed to operate with higher peak burnup, linear heat rate, and fast neutron fluence compared to typical LWRs. Therefore, assessment of fuel behavior is a critical task for its core performance. The fuel pellet radial power profile is calculated based on plutonium radial variation and edge peaking due to resonance absorption of neutrons. It is found that the edge power peak is much smaller than in typical LWRs due to the harder neutron spectrum. The oxygen potential directly affects fuel thermal conductivity and fission gas diffusivity. Plutonium migration towards the high temperature may potentially lead to power peaks at the central radial locations. The selected fuel thermal conductivity model for mixed oxides accounts for the oxygen-to-metal ratio variation, burnup effects due to fission product precipitates, radiation damage and porosity. In addition, Zircaloy-2 cladding corrosion/hydrogen pickup models in FRAPCON-3 have been updated to reflect accelerated corrosion/hydriding, due mainly to secondary particle precipitate dissolution. Based on experimental data, acceleration is assumed to occur above 10+26 n/m2 of fast neutron fluence (>1 MeV). Analysis of RBWR fuel was made together with neutron dose calculation using the reference power history. The neutron transport analysis shows that RBWR fuel fast fluence-to-volumetric heat generation ratio is approximately 80 % more than in typical LWRs. Initially, an analysis was performed with traditional Zircaloy-2 and reference mixed oxide fuel pellet with 95 % theoretical density. It was found that accelerated corrosion/hydriding may result at peak burnups as low as 30 MWd/kg. Furthermore, excessive fuel swelling may result in significant cladding strain and axial irradiation growth, which may lead to creep induced fracture as well as

  6. Oxidation Effect on Pool Boiling Heat Transfer in Atmospheric Saturated Water

    Energy Technology Data Exchange (ETDEWEB)

    Son, Hong Hyun; Jeong, Uiju; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of)

    2014-10-15

    During the hypothesized severe accidents, however, the modified nature of the oxidized outer surface of RPV may act as a significant heat transfer variable to achieve In-Vessel Retention through External Reactor Vessel Cooling (IVR-ERVC) strategy, which is the one of important mitigation strategies of severe accident to delay occurrence of critical heat flux (CHF). As well understood, the CHF is mainly affected by the two distinctive conditions classified to thermal hydraulic behavior of fluid system and surface characteristics. In this regard, a CHF test considering oxidation effect on the pool boiling heat transfer of the RPV outer surface has been proposed to evaluate realistic thermal margin of IVR-ERVC strategy. In this study, pool boiling heat transfer experiment was conducted under the condition of atmospheric saturated water. Oxidized surface characteristics were quantitatively evaluated with measurement of contact angle and roughness. In this study, oxide layer formation on the heated surface was investigated and experimentally simulated to find out its effect on the pool boiling CHF. Several SS316L substrates were oxidized in the corrosive environment under the condition of high temperature with different oxidation periods. Local pitting corrosion was observed on the heating surface in 5 days of short-term oxidation but a fully oxidized surface with somewhat uniform thickness, 1. Pool boiling heat transfer tests with the bare and oxidized heaters were conducted and major findings are summarized as follows: 1. Wettability in terms of the receding angle of the oxidized surface is enhanced regardless of the oxidation period. 2. Average roughness between the oxidized surfaces is almost the same in the range of nano-scale. 3. Effect of wettability and surface roughness on the CHF was negligible in the locally oxidized surface, which may be attributed to the presence of the disconnected porous channel. Unlike the local oxidation, fully oxidized surface shows

  7. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station

    International Nuclear Information System (INIS)

    This study presents the results of a comparison of a previous decommissioning cost study by Pacific Northwest Laboratory (PNL) and a recent decommissioning cost study of TLG Engineering, Inc., for the same commercial nuclear power reactor station. The purpose of this comparative analysis on the same plant is to determine the reasons why subsequent estimates for similar plants by others were significantly higher in cost and external occupational radiation exposure (ORE) than the PNL study. The primary purpose of the original study by PNL (NUREG/CR-0672) was to provide information on the available technology, the safety considerations, and the probable costs and ORE for the decommissioning of a large boiling water reactor (BWR) power station at the end of its operating life. This information was intended for use as background data and bases in the modification of existing regulations and in the development of new regulations pertaining to decommissioning activities. It was also intended for use by utilities in planning for the decommissioning of their nuclear power stations. The TLG study, initiated in 1987 and completed in 1989, was for the same plant, Washington Public Supply System's Unit 2 (WNP-2), that PNL used as its reference plant in its 1980 decommissioning study. Areas of agreement and disagreement are identified, and reasons for the areas of disagreement are discussed. 31 refs., 3 figs., 22 tabs

  8. Development of fast algorithms for estimating stress corrosion crack growth rate in sensitized stainless steels in boiling water reactor coolant environments

    International Nuclear Information System (INIS)

    A simplified method is proposed for modeling the chemistry and potential distribution in a stress corrosion crack in sensitized stainless steel in boiling water reactor (BWR) coolant environments. The model is based on an assumption that only those species that are present at the largest concentrations in the crevice determine the potential distribution down the crack. The advantage of this method is that it permits simplification of the mathematics and allows predictions to be made of the potential and concentration distributions without knowing various parameters, such as the equilibrium constants for homogeneous chemical reactions in the cavity and diffusion coefficients of species that are present at relatively low concentrations near the crack tip. In some important cases, analytical expressions can be obtained for the pH, potential near the crack tip, and crack propagation rate. The conditions for which the potential on the crack flanks and in the vicinity of the crack tip coincides with the free corrosion potential in the local environment, and hence for which there exists a balance between rates of the local anodic and cathodic partial processes, are determined. The impact of the potential drop in the external environment on the potential and concentration distributions down the crack and on the crack propagation rate is also investigated. Excellent agreement is obtained between calculated and measured crack growth rates

  9. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

  10. Experimental Research on Water Boiling Heat Transfer on Horizontal Copper Rod Surface at Sub-Atmospheric Pressure

    Directory of Open Access Journals (Sweden)

    Li-Hua Yu

    2015-09-01

    Full Text Available In recent years, water (R718 as a kind of natural refrigerant—which is environmentally-friendly, safe and cheap—has been reconsidered by scholars. The systems of using water as the refrigerant, such as water vapor compression refrigeration and heat pump systems run at sub-atmospheric pressure. So, the research on water boiling heat transfer at sub-atmospheric pressure has been an important issue. There are many research papers on the evaporation of water, but there is a lack of data on the characteristics at sub-atmospheric pressures, especially lower than 3 kPa (the saturation temperature is 24 °C. In this paper, the experimental research on water boiling heat transfer on a horizontal copper rod surface at 1.8–3.3 kPa is presented. Regression equations of the boiling heat transfer coefficient are obtained based on the experimental data, which are convenient for practical application.

  11. Halden Boiling Water Reactor. Plant Performance and Heavy-Water Management

    International Nuclear Information System (INIS)

    The Halden boiling heavy-water reactor, designed and built by the Norwegian Institutt for Atomenergi, has since June 1958 been operated as an international project. On its second charge the reactor was operated at power levels up to 25 MW and most of the time at a pressure of 28.5 kg/cm2. During the period from July 1964 to December 1966 the plant availability was close to 64% including shutdowns because of test fuel failures and loading/unloading of fuel. Disregarding such stops, the availability was close to 90%. The average burnup of the core is about 6200 MWd/t UO2 : the most highly exposed elements have reached 10000 MWd/t UO2. The transition temperature of the reactor tank has been followed closely. The results of the surveillance programme and the implication on the reactor operation are discussed. The reactor is located in a cave in a rock. Some experiences with such a containment are given. To locate failed test-fuel elements a fuel failure location system has been installed. A fission gas collection system has saved valuable reactor time during clean-up of the reactor system following test fuel failures. Apart from one incident with two of the control stations, the plant control and instrumentation systems have functioned satisfactorily. Two incidents with losses of 150 and 200 kg of heavy water have occurred. However, after improved methods for leakage detection had been developed, the losses have been kept better than 50 g/h . Since April 1962 the isotopic purity of the heavy water (14 t) has decreased from 99.75 to 99.62%. The tritium concentration is now slightly above 700 μC/cm3. This activity level has not created any serious operational or maintenance problems. An extensive series of water chemistry experiments has been performed to study the influence of various operating parameters on radiolytic gas formation. The main results of these experiments will be reported. Different materials such as mild steel, ferritic steel and aluminium have been

  12. Application of Spherical Copper Oxide (II) Water Nano-fluid as a Potential Coolant in a Boiling Annular Heat Exchanger

    OpenAIRE

    Nikkhah, V.; M. M. Sarafraz; F hormozi

    2015-01-01

    Convective boiling heat transfer coefficient of spherical CuO (II) nanoparticles dispersed in water is experimentally quantified inside the vertical heat exchanger. Influence of different operating parameters including applied heat and mass fluxes, sub-cooling temperature and concentration of nano-fluid on forced convection and nucleate boiling heat transfer mechanisms is experimentally investigated and briefly discussed. Results show that by increasing heat and mass fluxes, the heat trans...

  13. Experimental and Analytical Study of Lead-Bismuth-Water Direct Contact Boiling Two-Phase Flow

    Science.gov (United States)

    Novitrian; Dostal, Vaclav; Takahashi, Minoru

    The characteristics of lead-bismuth(Pb-Bi)-water boiling two-phase flow were investigated experimentally and analytically using a Pb-Bi-water direct contact boiling two-phase flow loop. Pb-Bi flow rates and void fraction were measured in a vertical circular tube at conditions of system pressure 7MPa, liquid metal temperature 460°C and injected water temperature 220°C. The drift-flux model with the assumption that bubble sizes were dependent on the fluid surface tension and the density ratio of Pb-Bi to steam-water mixture was chosen and modified by the best fit to the measured void fraction. Pb-Bi flow rates were analytically estimated using balance condition between buoyancy force and pressure losses, where the buoyancy force was calculated from void fraction estimated using the modified drift-flux model. The deviation of the analytical results of the flow rates from the experimental ones was less than 10%.

  14. Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  15. Production induced boiling and cold water entry in the Cerro Prieto geothermal reservoir indicated by chemical and physical measurements

    Energy Technology Data Exchange (ETDEWEB)

    Grant, M.A. (DSIR, Wellington, New Zealand); Truesdell, A.H.; Manon, A.

    1981-01-01

    Chemical and physical data suggest that the relatively shallow western part of the Cerro Prieto reservoir is bounded below by low permeability rocks, and above and at the sides by an interface with cooler water. There is no continuous permeability barrier around or immediately above the reservoir. Permeability within the reservoir is dominantly intergranular. Mixture with cooler water rather than boiling is the dominant cooling process in the natural state, and production causes displacement of hot water by cooler water, not by vapor. Local boiling occurs near most wells in response to pressure decreases, but no general vapor zone has formed.

  16. A method of simulating voids in experimental studies of boiling water reactors

    International Nuclear Information System (INIS)

    The coolant density in boiling water reactors may vary from 3 at pressures up to 1000 p.s.i. In order to study the effect of reduced water density on reactivity in unpressurized experimental systems, the effective water density is reduced by packing small beads of highly expanded polystyrene into the fuel clusters and flooding the interstices with water. Coolant densities of from 0.4 to 0.6 gm/cm3 may be produced with the introduction of only about 0.4 gm/cm3 of non-hydrogeneous material. This memorandum describes the production, properties and handling of polystyrene beads and the tests carried out to establish the validity of the technique. (author)

  17. Development of the neutron source evaluation method and predictor of SRM/SRNM count rate in BWR simulator

    International Nuclear Information System (INIS)

    The source range monitors (SRMs) and the start-up range neutron monitors (SRNMs) are important instruments from the BWR criticality safety viewpoints. There is a limitation of the minimum count rate (3cps) to guarantee the normality of the SRMs/SRNMs. After the long outage, this limitation is critical for the fuel shuffling due to the decay of the neutron sources in the fuel. The neutron source intensity evaluation method based on a micro burn-up model and the predictor function of the SRM/SRNM count rate are developed in AETNA01, GNF's three-dimensional neutronic-thermal hydraulic boiling water reactor (BWR) core simulator. These new functions are validated through the comparisons between operating BWR's measured data after shutdown and during shuffling. Through these comparisons, high accuracy of the SRM/SRNM count rate predictor of AETNA01 was presented. (author)

  18. The benefits of international cooperation via the Boiling Water Reactor Owners' Group (BWROG)

    International Nuclear Information System (INIS)

    The Boiling Water Reactors Owners' Group (BWROG) is an industry organization that was created in an effort to support common resolution of technical issues, address regulatory concerns, promote sharing of information and lessons learned among members, as well as to promote safety, minimize cost, and provide the proper forum for its members to address various specific issues. The BWROG is set up with an Executive Committee, responsible for overall organization performance, a General Committee responsible for day to day issues and operations, as well as numerous Technical Committees. BWROG represents almost 70 reactors worldwide and thousands years of operating experience

  19. Power distribution control within the scope of the advanced nuclear predictor for boiling water reactors

    International Nuclear Information System (INIS)

    In boiling water reactors the Advanced Nuclear Predictor (FNR) has proved to be a valuable tool in improving plant operating efficiency. The system is described in its main features and capabilities. As a logical extension, a power distribution control system has been developed, based on a reduced but accurate core model, which in itself can be used for fast prediction of core states. The system provides prediction of optimal operating strategies as well as on-line control, observing all constraints imposed on the permissible operating region. (orig.)

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

    International Nuclear Information System (INIS)

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

  1. Nuclear power plant with boiling water reactor VK-300 for district heating and electricity supply

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, Y.N.; Lisitza, F.D.; Romenkov, A.A.; Tokarev, Y.I. [RDIPE, Moscow (Russian Federation)

    1998-07-01

    The paper considers specific design features of a pressure vessel boiling water reactor with coolant natural circulation and three-step in-vessel steam separation (at draught tube outlet of the upcomer, within zone of overflow from the upcomer to downcomer and in cyclon-type separators). Design description and analytical study results are presented for the passive core cooling system in the case of loss of preferred power and rupture in primary circuit pipeline. Specific features of a primary containment (safeguard vessel) are given for an underground NPP sited in a rock ground. (author)

  2. Calculations of severe accident progression in the General Electric Simplified Boiling Water Reactor

    International Nuclear Information System (INIS)

    General Electric is designing a new nuclear power plant: the Simplified Boiling Water Reactor (SBWR). The SBWR is a passive plant in which the core cooling and decay heat removal safety systems are driven by gravity. To model the plant response to severe accidents, MAAP-SBWR, an advanced version of the Modular Accident Analysis Program (MAAP), has been developed. The main feature of the new code is a flexible containment model. The challenges in modeling the SBWR, the code structure and models, and a sample application to the SBWR are discussed

  3. Validation of HELIOS for calculations of experiments in the Halden Boiling Water Reactor

    International Nuclear Information System (INIS)

    The two dimensional transport code HELIOS has been used since the beginning of 1998 for the neutron physics calculations of experiments in the Halden Boiling Water Reactor (HBWR). Because a lot of measured data from these experiments are depending also on calculated values, it was necessary to validate HELIOS for these calculations. Therefore several experiments were re-calculated and it is shown that there is a good agreement between calculated and measured data. In some cases the effect of the calculated values on the measurements are shown, and this also shows that HELIOS gives reliable results

  4. Burnout in the boiling of water and freon-113 on tubes with annular fins

    International Nuclear Information System (INIS)

    This paper presents the results of numerical calculations of burnout heat flux associated with the boiling of Freon-113 and water on an annular fin of constant thickness which have been approximated by simple analytical relations. These are used to calculate the critical burnout parameters of tubes with an annular fin assembly. The calculated data may be used for the analysis of tubes with an annular fin assembly over a wide range of variation of the thermophysical properties of the material and geometrical parameters of the fin assembly

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

    International Nuclear Information System (INIS)

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

  6. Approaches to enhancing early hydrogen water chemistry for IGSCC mitigation during BWR startups

    International Nuclear Information System (INIS)

    Boiling Water Reactors (BWRs) have been injecting hydrogen into the reactor coolant via the feedwater system for the purpose of controlling primary system intergranular stress corrosion cracking (IGSCC) for over 30 years. However, plant design limitations prevent hydrogen injection until there is sufficient steam flow to support operation of the Steam Jet Air Ejector (SJAE) system, which typically occurs at greater than 5% power. The time from when the reactor coolant temperature is heated up to 200°F (93.3°C) until hydrogen injection starts is counted as time when IGSCC is not mitigated. Laboratory data show that crack growth rates peak at intermediate temperatures. To address this gap, Early Hydrogen Water Chemistry (EHWC) was developed by EPRI/BWRVIP for plants that use GE Hitachi NobleChem™ to lower the electrochemical corrosion potential (ECP) during early startup. A demonstration of EHWC performed at Peach Bottom 3 in October 2011 showed that sufficient hydrogen could be injected, while condenser vacuum was being maintained using the Mechanical Vacuum Pump (MVP), before steam flow was sufficient to place the SJAE system in service, to lower the ECP to a level indicative of IGSCC mitigation. While this demonstration successfully showed that sufficient hydrogen can be injected safely to mitigate IGSCC during early startup when temperature was between 360°F and 460°F (182 – 238°C), a reactor coolant hydrogen:oxidants molar ratio >2 was not achieved during the initial heatup when the temperature reached 200°F (93.3°C). This paper provides the results of a BWRVIP investigation of options to mitigate IGSCC at all temperatures above 200°F (93.3°C) for BWRs that apply noble metals. Options investigated for plants that have applied noble metal include 1) plant capabilities to achieve conditions at which IGSCC can be mitigated from 200°F (93.3°C) and above through the use of hydrogen alone and 2) the use of other agents, including hydrazine or methanol

  7. CORQUENCH: A model for gas sparging-enhanced, melt-water, film-boiling heat transfer

    International Nuclear Information System (INIS)

    In evaluation of severe-accident sequences for water-cooled nuclear reactors, molten core materials may be postulated to be released into the containment and accumulate on concrete. The heatup and decomposition of concrete is accompanied by the release of water vapor and carbon dioxide gases. Gases flowing through the melt upper surface can influence the rates of heat transfer to water overlying the melt. In particular, the gas flow through the interface can be envisioned to enhance the heat removal from the melt. A mechanistic model (CORQUENCH) has been developed to describe film-boiling heat transfer between a molten pool and an overlying coolant layer in the presence of sparging gas. The model favorably predicts the lead-Feron 11 data of Greene and Greene et al. for which the calculations indicate that area enhancement in the conduction heat transfer across the film is the predominant mechanism leading to augmentation in the heat flux as the gas velocity increases. Predictions for oxidic corium indicate a rapid increase in film-boiling heat flux as the gas velocity rises. The predominant mode of heat transfer for this case is radiation, and the increase in heat flux with gas velocity is primarily a result of interfacial area enhancement of the radiation component of the overall heat transfer coefficient. The CORQUENCH model has been incorporated into the MELTSPREAD-1 computer code6 for the analysis of transient spreading in containments

  8. Physical characteristics and antioxidant effect of polysaccharides extracted by boiling water and enzymolysis from Grifola frondosa.

    Science.gov (United States)

    Fan, Yina; Wu, Xiangyang; Zhang, Min; Zhao, Ting; Zhou, Ye; Han, Liang; Yang, Liuqing

    2011-06-01

    Grifola frondosa has been widely consumed in China and other Asian countries. Recent studies on G. frondosa have focused on the activities of polysaccharides extracted by water, and the activities of polysaccharides extracted by enzymolysis have not been studied. In this work, the relationship between the physical properties and antioxidant activity of polysaccharides extracted from G. frondosa by boiling water and enzymolysis was studied. Five polysaccharide extracts from the fruit body of G. frondosa were prepared by different extracting methods including boiling water, single enzyme enzymolysis with three different single enzymes (cellulose, pectinase, and pancreatin), and combined enzyme enzymolysis (cellulose:pectinase:pancreatin; 2:2:1). Characteristics such as the viscosity, Mw, polysaccharide content, protein content, infrared spectra, and antioxidant activities of the extracts were evaluated. The highest antioxidant activity was exhibited by the extracts prepared by combined enzyme extraction. The correlation analysis between antioxidant activity and polysaccharide content, protein content, Mw or viscosity indicated that the Mw had a more important role in antioxidant activity. Overall, the results indicate that the combined enzyme polysaccharide extracts can be developed as a new potential natural antioxidant. PMID:21458482

  9. Oscillate Boiling

    CERN Document Server

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

    2016-01-01

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

  10. Crack growth behavior of irradiated austenitic stainless steels in BWR environments

    International Nuclear Information System (INIS)

    Crack growth tests have been performed in boiling water reactor (BWR) environments on Types 304 and 316 stainless steel that were irradiated to fluence levels up to 2.0 x 1021 n cm-2 (E > 1 MeV) at approx. 288 degC in a helium environment. Two waveforms were used in the tests, slow/fast sawtooth and trapezoidal. The cyclic loading was done with rise times between 30 and 1000 s. At the longer rise times, the environmental contributions to the crack growth rate dominate. The trapezoidal waveform essentially represents constant load with periodic unloading and loading. The results indicate significant enhancement of crack growth rates of the irradiated steel in the BWR environment with normal water chemistry. The effects of fluence and hydrogen water chemistry are presented. (author)

  11. Heat transfer with high flux density between a wall and water, with local boiling at the wall

    International Nuclear Information System (INIS)

    The aim of the present study was to look for the relationship between the temperature of a hot wall and that of a cooling fluid under conditions where this fluid is subject to local boiling at the wall. Attention has been directed particularly to the aspect of the phenomenon in the transition region between the classical turbulent convection and the convection with local boiling. The tests were carried out with water at an average temperature of 50 deg. C circulating in an annular space, the heat flux deriving from the inside tube. The flux region used (300000 to 1 700 000 Kcal./h.m2) corresponds to a transmission by normal turbulent convection and by convection with the appearance of local boiling. The results obtained in the field of normal convection follow the classical laws, the results in the zone where local boiling appears are in correct agreement with the formulae proposed by other authors for higher heat fluxes. (author)

  12. Electrochemical response to hydrogen water chemistry at the J.A. FitzPatrick BWR

    International Nuclear Information System (INIS)

    It was the goal of the HWC campaign at the FitzPatrick BWR to determine the hydrogen injection rates required to mitigate IGSCC and IASCC in the reactor internals. Electrochemical sensors were installed at two elevations in one of the local power range monitors (LPRMs). In the summer of 1990 the HWC campaign was conducted. The feedwater hydrogen injection rate was varied from 12 to 90 standard cubic feet/minute (SCFM) and the ECPs from the sensors in the LPRM were measured. The relationship of hydrogen injection versus ECP was determined with specific emphasis on the injection rate required to decrease the ECP to -0.230 V(SHE) at each location in the LPRM. The LPRM lower position, equivalent to the outlet of the lower plenum, required three times more hydrogen injection than previously determined for the recirculation piping system to achieve -0.230 V(SHE). The upper position in the LPRM required far greater hydrogen injection rates to approach the protection potentials. Since completion of the FitzPatrick test, a program with similar objectives was conducted at an overseas BWR. It was found that in the high radiation environment of the core bypass newly designed platinum sensors performed quite adequately as reversible reference electrodes. These results provide a possible approach for protection of key reactor structurals with minimum hydrogen injection and low main steam line dose rates

  13. Effects of Subcooling on Liquid Wetted Area Fraction of Heater Surface during Nucleate Boiling of Water in a Pool

    Energy Technology Data Exchange (ETDEWEB)

    Park, Youngjae; Kim, Hyungdae [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-05-15

    Complex two-phase heat transfer phenomena such as nucleate boiling, critical heat flux, quenching and condensation affect the thermal performance of Light Water Reactors (LWRs) under normal operation and during transients/accidents. These phenomena are typically characterized by the presence of a liquid-vapor-solid contact line on the surface from/to which the heat is transferred. For example, in nucleate boiling, a significant fraction of the energy are transferred from a liquid meniscus underneath the boiling bubble through the evaporation and its dynamics. Thus, the measurement of liquid-vapor-solid contact line at the edge of the liquid meniscus and dynamics of liquid phase around the bubble are important to explain the boiling heat transfer. Therefore, various measurement techniques for liquid-vapor phase distribution on the heater surface have been developed, such as conductivity and optical probes, X-ray and γ-ray tomography, total reflection. Recently, a simple measurement technique to detect the liquid-vapor phase distribution has been developed using infrared thermography, named as DEPIcT or Detection of Phase by Infrared Thermography. In this study, effects of liquid subcooling on the phase distribution during nucleate boiling are investigated. DEPIcT technique was applied to measurement of wetted area fraction in pool boiling of water on the smooth surface with various liquid subcooling at atmospheric pressure.

  14. Effect of subcooling and wall thickness on pool boiling from downward-facing curved surfaces in water

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, M.S.; Glebov, A.G. [Univ. of New Mexico, Albuquerque, NM (United States)

    1995-09-01

    Quenching experiments were performed to investigate the effects of water subcooling and wall thickness on pool boiling from a downward-facing curved surface. Experiments used three copper sections of the same diameter (50.8 mm) and surface radius (148 mm), but different thickness (12.8, 20 and 30 mm). Local and average pool boiling curves were obtained at saturation and 5 K, 10 K, and 14 K subcooling. Water subcooling increased the maximum heat flux, but decreased the corresponding wall superheat. The minimum film boiling heat flux and the corresponding wall superheat, however, increased with increased subcooling. The maximum and minimum film boiling heat fluxes were independent of wall thickness above 20 mm and Biot Number > 0.8, indicating that boiling curves for the 20 and 30 thick sections were representative of quasi steady-state, but not those for the 12.8 mm thick section. When compared with that for a flat surface section of the same thickness, the data for the 12.8 mm thick section showed significant increases in both the maximum heat flux (from 0.21 to 0.41 MW/m{sup 2}) and the minimum film boiling heat flux (from 2 to 13 kW/m{sup 2}) and about 11.5 K and 60 K increase in the corresponding wall superheats, respectively.

  15. Effects of Subcooling on Liquid Wetted Area Fraction of Heater Surface during Nucleate Boiling of Water in a Pool

    International Nuclear Information System (INIS)

    Complex two-phase heat transfer phenomena such as nucleate boiling, critical heat flux, quenching and condensation affect the thermal performance of Light Water Reactors (LWRs) under normal operation and during transients/accidents. These phenomena are typically characterized by the presence of a liquid-vapor-solid contact line on the surface from/to which the heat is transferred. For example, in nucleate boiling, a significant fraction of the energy are transferred from a liquid meniscus underneath the boiling bubble through the evaporation and its dynamics. Thus, the measurement of liquid-vapor-solid contact line at the edge of the liquid meniscus and dynamics of liquid phase around the bubble are important to explain the boiling heat transfer. Therefore, various measurement techniques for liquid-vapor phase distribution on the heater surface have been developed, such as conductivity and optical probes, X-ray and γ-ray tomography, total reflection. Recently, a simple measurement technique to detect the liquid-vapor phase distribution has been developed using infrared thermography, named as DEPIcT or Detection of Phase by Infrared Thermography. In this study, effects of liquid subcooling on the phase distribution during nucleate boiling are investigated. DEPIcT technique was applied to measurement of wetted area fraction in pool boiling of water on the smooth surface with various liquid subcooling at atmospheric pressure

  16. Pool boiling heat transfer characteristics of vertical cylinder quenched by SiO{sub 2}-water nano-fluids

    Energy Technology Data Exchange (ETDEWEB)

    Abdurrahim, Bolukbasi; Dogan, Ciloglu [Department of Mechanical Engineering, Ataturk University, Erzurum (Turkey)

    2011-06-15

    This study includes an experimental investigation of the pool boiling heat transfer characteristics of a vertical cylinder quenched by SiO{sub 2}-water nano-fluids. The experiments are performed through a cylindrical rod, at saturated temperature and atmospheric pressure. As the coolant, pure water and SiO{sub 2}-water nano-fluid suspensions at four different concentrations (0.001, 0.01, 0.05 and 0.1 vol.%) are selected. The test specimen heated at high temperatures is plunged into cooling fluids at saturated conditions in a pool. The cooling curves are obtained via taking the temperature-time data of the specimen into account. The experimental results indicate that the pool film boiling heat transfer in nano-fluids is identical to that in pure water. However, during the repetition tests in nano-fluids with high concentrations, the film boiling region disappears, and the critical heat flux increases. In addition, the nucleate pool boiling heat transfer coefficient decreases compared with that of pure water, but a considerable decrease in nucleate pool boiling heat transfer is not observed with the repetition tests. A change in surface characteristics due to the deposition of nano-particles on the surface has a major effect on the quenching process. (authors)

  17. Benchmark for a 3D Monte Carlo boiling water reactor fluence computational package - MF3D

    International Nuclear Information System (INIS)

    A detailed three dimensional model of a quadrant of an operating BWR has been developed using MCNP to calculate flux spectrum and fluence levels at various locations in the reactor system. The calculational package, MF3D, was benchmarked against test data obtained over a complete fuel cycle of the host BWR. The test package included activation wires sensitive in both the fast and thermal ranges. Comparisons between the calculational results and test data are good to within ten percent, making the MF3D package an accurate tool for neutron and gamma fluence computation in BWR pressure vessel internals. (orig.)

  18. Characteristic of Local Boiling Heat Transfer of Ammonia / Water Binary Mixture on the Plate Type Evaporator

    Science.gov (United States)

    Okamoto, Akio; Arima, Hirofumi; Kim, Jeong-Hun; Akiyama, Hirokuni; Ikegami, Yasuyuki; Monde, Masanori

    Ocean thermal energy conversion (OTEC) and discharged thermal energy conversion (DTEC) are expected to be the next generation energy production systems. Both systems use a plate type evaporator, and ammonia or ammonia/water mixture as a working fluid. It is important to clarify heat transfer characteristic for designing efficient power generation systems. Measurements of local boiling heat transfer coefficients and visualization were performed for ammonia /water mixture (z = 0.9) on a vertical flat plate heat exchanger in a range of mass flux (7.5 - 15 kg/m2s), heat flux (15 - 23 kW/m2), and pressure (0.7 - 0.9 MPa). The result shows that in the case of ammonia /water mixture, the local heat transfer coefficients increase with an increase of vapor quality and mass flux, and decrease with an increase of heat flux, and the influence of the flow pattern on the local heat transfer coefficient is observed.

  19. New model of cobalt activity accumulation on stainless steel piping surfaces under boiling water reactor conditions

    International Nuclear Information System (INIS)

    A new technique for on-line measurement of corrosion amount and activity accumulation was developed. Cobalt activity accumulation tests were conducted under the normal water chemistry (NWC) condition (electrochemical corrosion potential (ECP): +0.15 V vs. SHE) and the hydrogen water chemistry (HWC) condition (ECP -0.30 V vs. SHE, -0.42 V vs. SHE) to evaluate cobalt activity accumulation under HWC conditions in boiling water reactors (BWRs). Total corrosion decreased and cobalt activity accumulation increased as ECP decreased. Experimental data were reproduced by a new model, in which cobalt activity deposits on oxide particle surfaces by absorption or replacement. This model estimated the cobalt activity accumulation under HWC conditions (ECP <-0.42 V vs. SHE) after 10000 h to be 12 times as large as that under NWC conditions (ECP +0.15 V vs. SHE). (author)

  20. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    International Nuclear Information System (INIS)

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise

  1. Valuation of BWR stability operating in natural circulation conditions

    International Nuclear Information System (INIS)

    Nowadays, the design of reactors having appropriate stability margins, the adoption of operating procedures avoiding possible unstable regions and the development of mitigation strategies to cope with inadvertent instability occurrences have strongly limited safety concerns in this regard. However, despite the obvious need for plant-specific Probabilistic Safety Assessment (PSA), BWR (boiling water reactor) transients of general interest can be identified and characterized as for example, overpressurization events, Large Break Loss of Coolant Accidents (LBLOCAs), feedwater temperature decrease, increase of core flow, main circulation pump flow rate increase, control rod withdrawal and others. Simulations of these complex scenarios have been improved by the utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes. In this work, the RELAP5/MOD3.3 thermal-hydraulic system code and the PARCS/2.4 3D neutron kinetic code have been adopted to predict the Peach Bottom BWR stability during recirculation pumps trip while the reactor is operating in a special region of power and core flow map. In the recirculation pump trip event, the stopping of the recirculation pumps causes a sharp decrease in the core flow, which generates a considerable negative reactivity insertion that tends to reduce power and, consequently, the amount of steam generated. The BWR reactor stability has been valuated during natural circulation conditions after the pumps trip event. The time evolution of the power and the related thermal-hydraulic parameters were investigated to analyse the behavior of the reactor for this special operation condition. (author)

  2. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  3. Flow boiling heat transfer of ammonia/water mixture in a plate heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Taboas, Francisco [Universidad de Cordoba, Campus de Rabanales, Edificio Leonardo da Vinci, 14014 Cordoba (Spain); Valles, Manel; Bourouis, Mahmoud; Coronas, Alberto [CREVER - Universitat Rovira i Virgili, Av. Paisos Catalans No. 26, 43007 Tarragona (Spain)

    2010-06-15

    The objective of this work is to contribute to the development of plate heat exchangers as desorbers for ammonia/water absorption refrigeration machines driven by waste heat or solar energy. In this study, saturated flow boiling heat transfer and the associated frictional pressure drop of ammonia/water mixture flowing in a vertical plate heat exchanger is experimentally investigated. Experimental data is presented to show the effects of heat flux between 20 and 50 kW m{sup -2}, mass flux between 70 and 140 kg m{sup -2} s{sup -1}, mean vapour quality from 0.0 to 0.22 and pressure between 7 and 15 bar, for ammonia concentration between 0.42 and 0.62. The results show that for the selected operating conditions, the boiling heat transfer coefficient is highly dependent on the mass flux, whereas the influence of heat flux and pressure are negligible mainly at higher vapour qualities. The pressure drop increases with increasing mass flux and quality. However, the pressure drop is independent of the imposed heat flux. (author)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Cheng, Ling

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

  6. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    The present work presents the modeling and simulation of the aspersion system to low pressure of reactor of the nuclear power plant of Laguna Verde using the nuclear code RELAP/SCDAP. The objective of the emergency systems inside a nuclear reactor is the cooling of the core, nor caring the performance of any other emergency system in the case of an accident design base for coolant loss. To obtain a simulation of the system is necessary to have a model based on their main components, pipes, pumps, valves, etc. This article describes the model for the simulation of the main line and the test line for the HPCS. At the moment we have the simulation of the reactor vessel and their systems associated to the nuclear power plant of Laguna Verde, this work will allow to associate the emergency system model LPCS to the vessel model. The simulation of the vessel and the emergency systems will allow knowing the behavior of the reactor in the stage of the coolant loos, giving the possibility to analyze diverse scenarios. The general model will provide an auxiliary tool for the training in classroom and at distance in the operation of nuclear power plants. (Author)

  7. An assessment of BWR [boiling water reactor] Mark III containment challenges, failure modes, and potential improvements in performance

    International Nuclear Information System (INIS)

    This report describes risk-significant challenges posed to Mark III containment systems by severe accidents as identified for Grand Gulf. Design similarities and differences between the Mark III plants that are important to containment performance are summarized. The accident sequences responsible for the challenges and the postulated containment failure modes associated with each challenge are identified and described. Improvements are discussed that have the potential either to prevent or delay containment failure, or to mitigate the offsite consequences of a fission product release. For each of these potential improvements, a qualitative analysis is provided. A limited quantitative risk analysis is provided for selected potential improvements. 21 refs., 5 figs., 46 tabs

  8. Conceptual design and safety characteristics of the natural circulation boiling water reactor HSBWR-600

    International Nuclear Information System (INIS)

    The HSBWR (Hitachi Small BWR) with a rated capacity of 600 MW electricity has been conceptually designed. The components and systems are simplified by adopting natural circulation and the passive ECCS, and eliminating steam separators. The volume of the reactor building is about 50% of that for current BWRs with the same rated capacity, and the construction period is 32-36 months until commercial operation. The major safety systems are: (1) an accumulated water injection system as an ECCS; (2) an outer pool, which stands outside of the steel primary containment vessel, as a long term cooling system after LOCAs; and (3) a steam driven reactor core isolation cooling system for high pressure water injection. The grace period is one day for core cooling and 3 days for the containment vessel heat removal. The infinite grace period for core cooling is also available as an option. LOCA analysis showed that the core will always be covered by a two-phase mixture, resulting in no core heat-up. The fundamental experiments and analyses showed sufficient capability of the outer pool for long term heat removal. (author). 12 refs, 17 figs, 3 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-11-01

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

  10. Investigations on heat transfer enhancement in pool boiling with water-CuO nano-fluids

    Science.gov (United States)

    Hegde, Ramakrishna N.; Rao, Shrikantha S.; Reddy, R. P.

    2012-04-01

    The main focus of the present work is to investigate Critical Heat Flux (CHF) enhancement using CuO nanofluid relative to CHF of pure water. To estimate the effect of nanoparticles on the CHF, pool boiling CHF values were measured for various volume concentrations of CuO nanofluid and compared with pure water. CHF enhancement of 130% was recorded at 0.2 % by volume of CuO nano-fluids. Surface roughness of the heater surface exposed to three measured heating cycles indicated surface modifications at different volume concentrations of nanofluid. SEM image of the heater surface revealed porous layer build up, which is thought to be the reason for CHF enhancement.

  11. Nickel Catalyzed Conversion of Cyclohexanol into Cyclohexylamine in Water and Low Boiling Point Solvents

    Directory of Open Access Journals (Sweden)

    Yunfei Qi

    2016-04-01

    Full Text Available Nickel is found to demonstrate high performance in the amination of cyclohexanol into cyclohexylamine in water and two solvents with low boiling points: tetrahydrofuran and cyclohexane. Three catalysts, Raney Ni, Ni/Al2O3 and Ni/C, were investigated and it is found that the base, hydrogen, the solvents and the support will affect the activity of the catalyst. In water, all the three catalysts achieved over 85% conversion and 90% cyclohexylamine selectivity in the presence of base and hydrogen at a high temperature. In tetrahydrofuran and cyclohexane, Ni/Al2O3 exhibits better activity than Ni/C under optimal conditions. Ni/C was stable during recycling in aqueous ammonia, while Ni/Al2O3 was not due to the formation of AlO(OH.

  12. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  13. Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

    International Nuclear Information System (INIS)

    The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced power. However, the effect of reduced cycling frequency of the SRV due to the IC, and the time lag between the events, are comparable to that for forced convection

  14. Transient pool boiling heat transfer due to increasing heat inputs in subcooled water at high pressures

    Energy Technology Data Exchange (ETDEWEB)

    Fukuda, K. [Kobe Univ. of Mercantile Marine (Japan); Shiotsu, M.; Sakurai, A. [Kyoto Univ. (Japan)

    1995-09-01

    Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q{sub max}, on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q{sub o}e{sup t/T}, with periods, {tau}, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q{sub max}. Two main mechanisms of q{sub max} exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q{sub max} for long period range belonging to the former mechanism becomes longer and the q{sub max}mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q{sub max} for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling.

  15. Numerical investigation of water-based nanofluid subcooled flow boiling by three-phase Euler-Euler, Euler-Lagrange approach

    Science.gov (United States)

    Valizadeh, Ziba; Shams, Mehrzad

    2016-08-01

    A numerical scheme for simulating the subcooled flow boiling of water and water-based nanofluids was developed. At first, subcooled flow boiling of water was simulated by the Eulerian multiphase scheme. Then the simulation results were compared with previous experimental data and a good agreement was observed. In the next step, subcooled flow boiling of water-based nanofluid was modeled. In the previous studies in this field, the nanofluid assumed as a homogeneous liquid and the two-phase scheme was used to simulate its boiling. In the present study, a new scheme was used to model the nanofluid boiling. In this scheme, to model the nanofluid flow boiling, three phases, water, vapor and nanoparticles were considered. The Eulerian-Eulerian approach was used for modeling water-vapor interphase and Eulerian-Lagrangian scheme was selected to observe water-nanoparticle interphase behavior. The results from the nanofluid boiling modeling were validated with an experimental investigation. The results of the present work and experimental data were consistent. The addition of 0.0935 % volume fraction of nanoparticles in pure liquid boiling flow increases the vapor volume fraction at the outlet almost by 40.7 %. The results show the three-phase model is a good approach to simulate the nanofluid boiling flow.

  16. Applied methods for mitigation of damage by stress corrosion in BWR type reactors

    International Nuclear Information System (INIS)

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  17. Heat transfer and critical heat flux of subcooled water flow boiling in a short horizontal tube

    International Nuclear Information System (INIS)

    The steady-state turbulent heat transfer (THT) due to exponentially increasing heat inputs with various exponential periods (Q=Q0exp(t/τ), τ=6.55 to 21.81 s) were systematically measured with the flow velocities, u, of 4.15, 7.05, 10.07 and 13.50 m/s by an experimental water loop flow. Measurements were made on a 6 mm inner diameter, a 59.2 mm effective length and a 0.4 mm thickness of HORIZONTAL Platinum (Pt) circular test tube. The relation between the steady-state turbulent heat transfer and the flow velocity were clarified. The steady state nucleate boiling heat transfer (NBHT) and the steady state critical heat fluxes (CHFs) of the subcooled water flow boiling for HORIZONTAL SUS304 circular test tube were systematically measured with the flow velocities (u=3.94 to 13.86 m/s), the inlet subcoolings (ΔTsub,in=81.30 to 147.94 K), the inlet pressures (Pin=786.29 to 960.93 kPa) and the increasing heat input (Q0 exp(t/τ), τ=8.36 s). The HORIZONTAL SUS304 test tube of inner diameter (d=6 mm), heated length (L=59.4 mm), effective length (Leff=48.4 mm), L/d (=9.9), Leff/d (=8.06) and wall thickness (δ=0.5 mm) with surface roughness (Ra=3.89 μm) was used in this work. The NBHT and the steady state CHFs of the subcooled water flow boiling for the HORIZONTAL SUS304 test tube were clarified at the flow velocities u ranging from 3.94 to 13.86 m/s. The steady-state THT data, the NBHT ones and the steady state CHF ones were compared with the values calculated by authors' THT correlation, their NBHT ones and their transient CHF ones against outlet and inlet subcoolings based on the experimental data for the VERTICAL circular test tubes with the flow velocities u ranging from 4.0 to 42.4 m/s. The influences of test tube orientation on the THT, the NBHT and the subcooled flow boiling CHF are investigated into details and the widely and precisely predictable correlations of the THT, the NBHT and the transient CHFs against outlet and inlet subcoolings in a short

  18. Measurement of wetted area fraction in subcooled pool boiling of water using infrared thermography

    International Nuclear Information System (INIS)

    The wetted area fraction in subcooled pool boiling of water at atmospheric pressure is measured using the DEPIcT (DEtection of Phase by Infrared Thermography) technique. DEPIcT exploits the contrast in infrared (IR) light emissions between wet and dry areas on the surface of an IR-transparent heater to visualize the instantaneous distribution of the liquid and gas phases in contact with the heater surface. In this paper time-averaged wetted area fraction data in nucleate boiling are reported as functions of heat flux (from 30% up to 100% of the Critical Heat Flux) and subcooling (ΔTsub = 0, 5, 10, 30 and 50 °C). The results show that the wetted area fraction monotonically decreases with increasing heat flux and increases with increasing subcooling: both trends are expected. The range of time-averaged wetted area fractions is from 90%, at low heat flux and high subcooling, to 50% at high heat flux (right before CHF) and low subcooling. It is also shown that the dry areas are periodically rewetted by liquid sloshing on the surface at any subcooling and heat flux; however, the dry areas expand irreversibly at CHF

  19. Conceptual design of a self-sustainable pressurized water reactor with boiling channels

    International Nuclear Information System (INIS)

    Parametric studies have been performed on a seed-blanket Th-U233 fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts required substantial reduction of the core power density in order to operate under nominal PWR system conditions. Boiling flow regime in the seed area allows better heat removal, which in turn, may potentially allow increasing the power density of the core. In addition, the reduced moderation improves the breeding performance. A 2-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to 104 W/cc, created a map of designs with their corresponding fissile inventory ratio (FIR) values. It was found that several options have the potential to achieve the main objective - a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. (author)

  20. Performance of Charcoal Cookstoves for Haiti Part 1: Results from the Water Boiling Test

    Energy Technology Data Exchange (ETDEWEB)

    Booker, Kayje; Han, Tae Won; Granderson, Jessica; Jones, Jennifer; Lsk, Kathleen; Yang, Nina; Gadgil, Ashok

    2011-06-01

    In April 2010, a team of scientists and engineers from Lawrence Berkeley National Lab (LBNL) and UC Berkeley, with support from the Darfur Stoves Project (DSP), undertook a fact-finding mission to Haiti in order to assess needs and opportunities for cookstove intervention. Based on data collected from informal interviews with Haitians and NGOs, the team, Scott Sadlon, Robert Cheng, and Kayje Booker, identified and recommended stove testing and comparison as a high priority need that could be filled by LBNL. In response to that recommendation, five charcoal stoves were tested at the LBNL stove testing facility using a modified form of version 3 of the Shell Foundation Household Energy Project Water Boiling Test (WBT). The original protocol is available online. Stoves were tested for time to boil, thermal efficiency, specific fuel consumption, and emissions of CO, CO{sub 2}, and the ratio of CO/CO{sub 2}. In addition, Haitian user feedback and field observations over a subset of the stoves were combined with the experiences of the laboratory testing technicians to evaluate the usability of the stoves and their appropriateness for Haitian cooking. The laboratory results from emissions and efficiency testing and conclusions regarding usability of the stoves are presented in this report.