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Sample records for bwr boiling water

  1. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  2. Boiling water system of nuclear power plants (BWR)

    International Nuclear Information System (INIS)

    Martias Nurdin

    1975-01-01

    About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

  3. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  4. Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Peng, C.M.; Maly, J.

    1988-01-01

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

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

    International Nuclear Information System (INIS)

    Ramirez G, C.; Chavez M, C.

    2012-10-01

    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)

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

    International Nuclear Information System (INIS)

    Behrooz, A.

    2008-01-01

    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

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

    International Nuclear Information System (INIS)

    Fuentes C, P.

    2003-01-01

    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 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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    1987-11-01

    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

  11. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

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

    International Nuclear Information System (INIS)

    Bottimore, R.R.

    1980-12-01

    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

  13. A real-time BWR [boiling water reactor] stability measurement system

    International Nuclear Information System (INIS)

    March-Leuba, J.; King, W.T.

    1987-01-01

    This paper describes the characteristics of a portable, real-time system used for nonperturbational measurements of stability in boiling water reactors. The algorithm used in this system estimates the closed-loop asymptotic decay ratio using only the naturally occurring neutron noise and it is based on the univariate autoregressive methodology

  14. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

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

    1989-06-01

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

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

    International Nuclear Information System (INIS)

    Holman, G.S.

    1989-12-01

    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

  16. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    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)

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

    International Nuclear Information System (INIS)

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

    1992-10-01

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

  18. Stress corrosion cracking of L-grade stainless steels in boiling water reactor (BWR) plants

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Fukuda, Toshihiko; Yamashita, Hironobu

    2004-01-01

    L-grade stainless steels as 316NG, SUS316L and SUS304L have been used for the BWR reactor internals and re-circulation pipes as SCC resistant materials. However, SCC of the L-grade material components were reported recently in many Japanese BWR plants. The detail investigation of the components showed the fabrication process such as welding, machining and surface finishing strongly affected SCC occurrence. In this paper, research results of SCC of L-grade stainless steels, metallurgical investigation of core shrouds and re-circulation pipings, and features of SCC morphology were introduced. Besides, the structural integrity of components with SCC, countermeasures for SCC and future R and D planning were introduced. (author)

  19. 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

  20. The BWR [Boiling Water Reactor] Emergency Operating Procedures Tracking System (EOPTS): Evaluation by control-room operating crews

    International Nuclear Information System (INIS)

    Spurgin, A.J.; Orvis, D.D.; Spurgin, J.P.; Luna, C.J.

    1990-05-01

    This report presents the results of a project sponsored by the Electric Power Research Institute (EPRI) and Taiwan Power Company (TPC) and conducted by APG and TPC to perform evaluation of the Emergency Operating Procedures Tracking System (EOPTS). The EOPTS is an expert system employing artificial intelligence techniques developed by EPRI for Boiling Water Reactor (BWR) plants based on emergency operating procedures (EOPs). EOPTS is a computerized decision aid used to assist plant operators in efficient and reliable use of EOPs. The main objective of this project was to evaluate the EOPTS and determine how an operator aid of this type could noticeably improve the response time and the reliability of control room crews to multi-failure scenarios. A secondary objective was to collect data on how crew performance was affected. Experiments results indicate that the EOPTS measurably improves crew performance over crews using the EOP flow charts. Time-comparison measurements indicate that crews using the EOPTS perform required actions more quickly than do those using the flowcharts. The results indicate that crews using the EOPTS are not only faster and more consistent in their actions but make fewer errors. In addition, they have a higher likelihood of recovering from the errors that they do make. Use of the EOPTS in the control room should result in faster termination and mitigation of accidents and reduced risk of power plant operations. Recommendations are made towards possible applications of the EOPTS to operator training and evaluation, and for the applicability of the evaluation methodology developed for this project to the evaluation of similar operator aides. 17 refs., 14 figs., 14 tabs

  1. The physico-chemical 131I species in the exhaust air of a boiling water reactor (BWR 4)

    International Nuclear Information System (INIS)

    Deuber, H.

    1982-12-01

    In a German boiling water reactor, the physico-chemical 131 I species were determined in the plant exhaust and in the individual exhausts during 12 months. These measurements aimed in particular at determining the percentage and the source of the radiologically decisive elemental 131 I released to the environment. The retention of the 131 I species by iodine filters was also investigated. On an average, 45% of the 131 I discharged with the plant exhaust consisted of elemental iodine. This was largely released with the exhaust from the reactor building and from the turbine building. The other 55% consisted almost entirely of organic I. (orig./HP) [de

  2. The physico-chemical I-131 species in the exhaust air of a boiling water reactor (BWR 5)

    International Nuclear Information System (INIS)

    Deuber, H.

    1984-02-01

    In a German boiling water reactor, the pysico-chemical I-131 species were determined in the plant exhaust and in the individual exhausts during four months. These measurements aimed in particular at determining the percentage and the source of the radiologically decisive elemental I-131 released to the environment. On an average 13% of the I-131 discharged with the plant exhaust consisted of elemental iodine. This was largely released with the exhausts from the reactor building and from the turbine building. The main component was organic-bound I. (orig./HP) [de

  3. Dynamics of a BWR with inclusion of boiling nonlinearity, clad temperature and void-dependent core power removal: Stability and bifurcation characteristics of advanced heavy water reactor (AHWR)

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

    Highlights: • Simplified models with inclusion of the clad temperature are considered. • Boiling nonlinearity and core power removal have been modeled. • Method of multiple time scales has been used for nonlinear analysis to get the nature and amplitude of oscillations. • Incorporation of modeling complexities enhances the stability of system. • We find that reactors with higher nominal power are more desirable from the point of view of global stability. - Abstract: We study the effect of including boiling nonlinearity, clad temperature and void-dependent power removal from the primary loop in the mathematical modeling of a boiling water reactor (BWR) on its dynamic characteristics. The advanced heavy water reactor (AHWR) is taken as a case study. Towards this end, we have analyzed two different simplified models with different handling of the clad temperature. Each of these models has the necessary modifications pertaining to boiling nonlinearity and power removal from the primary loop. These simplified models incorporate the neutronics and thermal–hydraulic coupling. The effect of successive changes in the modeling assumptions on the linear stability of the reactor has been studied and we find that incorporation of each of these complexities in the model increases the stable operating region of the reactor. Further, the method of multiple time scales (MMTS) is exploited to carry out the nonlinear analysis with a view to predict the bifurcation characteristics of the reactor. Both subcritical and supercritical Hopf bifurcations are present in each model depending on the choice of operating parameters. These analytical observations from MMTS have been verified against numerical simulations. A parametric study on the effect of changing the nominal reactor power on the regions in the parametric space of void coefficient of reactivity and fuel temperature coefficient of reactivity with sub- and super-critical Hopf bifurcations has been performed for all

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-01-01

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

  5. 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)

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

    International Nuclear Information System (INIS)

    Hazelton, W.S.; Koo, W.H.

    1988-01-01

    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

  7. Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fifth Workshop (BWR-TT5)

    International Nuclear Information System (INIS)

    2003-01-01

    The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fifth workshop was to discuss the results from Phase III (best

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

    International Nuclear Information System (INIS)

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

    1990-07-01

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

  9. Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fourth Workshop (BWR-TT4)

    International Nuclear Information System (INIS)

    2002-01-01

    The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fourth workshop was to present and discuss final results of

  10. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

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

    2012-09-01

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

  11. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    Stancavage, P.

    1991-01-01

    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)

  12. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    Yoshio Matsuo

    1987-01-01

    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)

  13. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  14. Singular system analysis of the Local Power Range Monitor (LPRM) readings of a Boiling Water Reactor (BWR) in an unstable event

    International Nuclear Information System (INIS)

    Ginestar Peiro, D.; Verdu, G.; Miro, R.

    2006-01-01

    Singular system analysis is a successful technique to separate oscillating components from a given signal. A methodology is proposed to apply this technique to the signals obtained from the LPRMs of a boiling water reactor core and extract the contributions of the in-phase oscillation and the out-of-phase oscillations from the LPRM readings during an unstable event. This methodology has been validated with synthetic signals and simulations of in-phase and out-of-phase oscillations of the Leibstadt reactor. Finally, one case of Ringhals I Stability Benchmark has been analysed. (author)

  15. A model for oxidizing species concentrations in boiling water reactors

    International Nuclear Information System (INIS)

    Sun, B.; Chexal, B.; Pathania, R.; Chun, J.; Ballinger, R.; Abdollahian, D.

    1993-01-01

    To evaluate and control the intergranular stress corrosion cracking of boiling water reactor (BWR) vessel internal components requires knowledge of the concentration of oxidizing species that affects the electrochemical potentials in various regions of a BWR. In a BWR flow circuit, as water flows through the radiation field, the radiolysis process and chemical reactions lead to the production of species such as oxygen, hydrogen, and hydrogen peroxide. Since chemistry measurements are difficult inside BWRs, analytical tools have been developed by Ruiz and Lin, Ibe and Uchida and Chun and Ballinger for estimating the concentration of species that provide the necessary input for water chemistry control and material protection

  16. Analysis of a main steam isolation valve closure anticipated transient without scram in a boiling water reactor

    International Nuclear Information System (INIS)

    Liaw, T.J.; Pan, C.; Chen, G.S.

    1989-01-01

    Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main stream isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4

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

    International Nuclear Information System (INIS)

    March-Leuba, J.; Perez, R.B.; Cacuci, D.G.

    1984-01-01

    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)

  18. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

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

    1990-03-01

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

  19. Future directions in boiling water reactor design

    International Nuclear Information System (INIS)

    Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

    1987-01-01

    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 design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; 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 trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

  20. Water level measurement uncertainty during BWR instability

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  1. Nonlinear dynamics of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1983-01-01

    Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

  2. 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)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C.

    2012-10-01

    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)

  3. 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)

    Bierschbach, M.C.

    1994-12-01

    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

  4. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  5. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    OpenAIRE

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  6. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    Luk, K.H.

    1993-09-01

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

  7. Boiling water reactor

    International Nuclear Information System (INIS)

    Eckardt, B.; Queiser, H.

    1986-01-01

    On BWR's, the turbine condenser is connected to a waste gas system with redundancy, which has the following in sequence in each of two parallel circuits: a steam jet pump, a valve, a recombiner with steam heating, a condenser and a gas outlet. Steam and air are fed in a ratio of 1:10 or more between the valve and the recombiner. A steam jet pump is used for this purpose. (orig./HP) [de

  8. Effective void fraction for a BWR assembly with boiling in the bypass region

    International Nuclear Information System (INIS)

    Galperin, A.; Segev, M.; Knoglinger, E.

    1991-09-01

    Average BWR assembly cross-sections for nominal conditions, namely for zero bypass void, can be utilised in the analysis of transient conditions with boiling in the bypass. A model is developed to yield an effective channel void for such conditions. The use of this void in conjunction with the 'nominal conditions' cross section library approximately preserves the assembly K-infinity corresponding to the true channel and bypass voids. The effective void is an augmentation of the actual channel void. The augment is proportional to the bypass-to-channel volume ratio, to the bypass void, and to a weight W which is introduced to quantify the fact that a water molecule in the bypass has a different assembly criticality worth than one in the channel. The formula developed is superior to the practice of choosing W=1, namely a simple, non-weighted, transfer of water from channel to bypass. The use of this approximate effective channel void reproduces actual K-infinity values of assemblies to better than 5 mk, whereas the use of a simple model sometimes misspredicts the assembly K-infinity by 40 mK. The effective void model cannot handle cases in which both channel and bypass void value are high, simply because then the effective void α ch eff becomes meaningless. A method to treat the α eff >1 domain is developed by which corrections to cross sections are provided. Such corrections are synthesised as functions of the assembly parameters. (author) figs., tabs., refs

  9. Analysis of water hammer in control rod drive systems of boiling water reactor nuclear power plants

    International Nuclear Information System (INIS)

    Safwat, H.H.; Arastu, A.H.; Lau, S.

    1983-01-01

    The method of characteristics is applied to analyze water hammer in BWR (Boiling Water Reactor) Control Rod Drive (CRD) Systems following fast opening of scram valves. The modelling of the CRD mechanism is presented. Numerical predictions are compared to experimental data. (author)

  10. Power distribution effects on boiling water reactor stability

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.

    1989-01-01

    The work presented in this paper deals with the effects of spatial power distributions on the stability of boiling water reactors (BWRs). It is shown that a conservative power distribution exists for which the stability is minimal. These results are relevant because they imply that bounding stability calculations are possible and, thus, a worst-possible scenario may be defined for a particular BWR geometry. These bounding calculations may, then, be used to determine the maximum expected limit-cycle peak powers

  11. 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)

    Vargas O, D.; Chavez M, C.

    2012-10-01

    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. Advanced boiling water reactors for the 90's and beyond

    International Nuclear Information System (INIS)

    Rao, A.S.; Sawyer, C.D.; Qurik, J.F.; McCandless, R.J.

    1990-01-01

    This paper discusses how the advanced boiling water reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability and reduced occupational exposure and radwaste. International cooperative efforts are also under way aimed at development of a simplified BWR employing natural circulation and passive safety systems. The SBWR conceptual design is complete. This BWR concept shows technical and economic promise. The SBWR program is aimed at providing a U.S. NRC certified design in an investor-ready state by 1995. With its short construction schedule, the 600 MWe SBWR will provide an option for commercial operation worldwide by the mid-to-late 1990s

  13. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2005-01-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermalhydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable. (author)

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

    International Nuclear Information System (INIS)

    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

  15. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

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

  16. When water does not boil at the boiling point.

    Science.gov (United States)

    Chang, Hasok

    2007-03-01

    Every schoolchild learns that, under standard pressure, pure water always boils at 100 degrees C. Except that it does not. By the late 18th century, pioneering scientists had already discovered great variations in the boiling temperature of water under fixed pressure. So, why have most of us been taught that the boiling point of water is constant? And, if it is not constant, how can it be used as a 'fixed point' for the calibration of thermometers? History of science has the answers.

  17. Review of boiling water reactor small break loss of coolant accidents

    International Nuclear Information System (INIS)

    Gururaj, P.M.; Dua, S.S.; Rao, A.S.

    1981-01-01

    This paper presents a review of the analytical and the experimental work performed by the General Electric Company to determine the performance of boiling water reactors (BWR) following postulated small break accidents (SBA). This review paper addresses the following issues: (1) the response of the BWR following small loss of inventory events; (2) methods of analysis and their justification; (3) necessity, if any, of operator action and the length of time available in which such action can be performed; and (4) operator interface following the SBA event. The results from these SBA studies for different BWR product lines show that even with the multiple system failures assumed, the BWR can successfully withstand an SBA. For a typical BWR/6, it takes the failure of 13 water delivery pumps to cause any significant core heatup. The only operator actions determined to be necessary are simple ones and ample time is available to the operator to perform these actions, if needed

  18. A review of boiling water reactor water chemistry: Science, technology, and performance

    International Nuclear Information System (INIS)

    Fox, M.J.

    1989-02-01

    Boiling water reactor (BWR) water chemistry (science, technology, and performance) has been reviewed with an emphasis on the relationships between BWR water quality and corrosion fuel performance, and radiation buildup. A comparison of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.56, the Boiling Water Reactor Owners Group (BWROG) Water Chemistry Guidelines, and Plant Technical Specifications showed that the BWROG Guidelines are more stringent than the NRC Regulatory Guide, which is almost identical to Plant Technical Specifications. Plant performance with respect to BWR water chemistry has shown dramatic improvements in recent years. Up until 1979 BWRs experienced an average of 3.0 water chemistry incidents per reactor-year. Since 1979 the water chemistry technical specifications have been violated an average of only 0.2 times per reactor-year, with the most recent data from 1986-1987 showing only 0.05 violations per reactor-year. The data clearly demonstrate the industry-wide commitment to improving water quality in BWRs. In addition to improving water quality, domestic BWRs are beginning to switch to hydrogen water chemistry (HWC), a remedy for intergranular stress corrosion cracking. Three domestic BWRs are presently operating on HWC, and fourteen more have either performed HWC mini tests or are in various stages of HWC implementation. This report includes a detailed review of HWC science and technology as well as areas in which further research on BWR chemistry may be needed. 43 refs., 30 figs., 8 tabs

  19. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  20. Water chemistry in boiling water reactors - A Leibstadt-specific overview

    International Nuclear Information System (INIS)

    Sarott, F.-A.

    2005-01-01

    The boiling water reactor (BWR) consists of two main water circuits: the water-steam cycle and the main cooling water system. In the introduction, the goals and tasks of the BWR plant chemistry are described. The most important objectives are the prevention of system degradation by corrosion and the minimisation of radiation fields. Then a short description of the BWR operation principle, including the water steam cycle, the transport of various impurities by the steam, removing impurities from the condensate, the reactor water clean-up system, the balance of plant and the main cooling water system, is given. Subsequently, the focus is set on the water-steam cycle chemistry. In order to fulfil the somewhat contradictory requirements, the chemical parameters must be well balanced. This is achieved by the water chemistry control method called 'normal water chemistry'. Other additional methods are used for the solution to different problems. The 'zinc addition method' is applied to reduce high radiation levels around the recirculation loops. The 'hydrogen water chemistry method' and the 'noble metal chemical addition method' are used to protect the reactor core components and piping made of stainless steel against stress corrosion cracking. This phenomenon has been observed for about 40 years and is partly due to the strong oxidising conditions in the BWR water. Both mitigation methods are used by the majority of the BWR plants all over the world (including the two Swiss NPPs Muehleberg and Leibstadt). (author)

  1. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

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

    International Nuclear Information System (INIS)

    Ishida, Naoyuki; Utsuno, Hideaki; Kasahara, Fumio

    2003-01-01

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

  3. Boiling water reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

    1975-01-01

    Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

  4. Approximation model of three-dimensional power distribution in boiling water reactor using neural networks

    International Nuclear Information System (INIS)

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2001-01-01

    Fast and accurate prediction of three-dimensional (3D) power distribution is essential in a boiling water reactor (BWR). The prediction method of 3D power distribution in BWR is developed using the neural network. Application of the neural network starts with selecting the learning algorithm. In the proposed method, we use the learning algorithms based on a class of Quasi-Newton optimization techniques called Self-Scaling Variable Metric (SSVM) methods. Prediction studies were done for a core of actual BWR plant with octant symmetry. Compared to classical Quasi-Newton methods, it is shown that the SSVM method reduces the number of iterations in the learning mode. The results of prediction demonstrate that the neural network can predict 3D power distribution of BWR reasonably well. The proposed method will be very useful for BWR loading pattern optimization problems where 3D power distribution for a huge number of loading patterns (LPs) must be performed. (author)

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

    International Nuclear Information System (INIS)

    Alder, H.P.; Schenker, E.

    1993-01-01

    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

  6. Technical report on operating experience with boiling water reactor offgas systems

    International Nuclear Information System (INIS)

    Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

    1978-03-01

    Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

  7. Experiences in stability testing of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Otaduy, P.J.

    1986-01-01

    The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

  8. An appraisal of subcooled boiling and slip ratio from measurements made in Lingen BWR

    International Nuclear Information System (INIS)

    Nash, G.

    1977-08-01

    Measurements of steam bubble velocities and voidage have been made in the relatively small Core B of Lingen BWR. The results of axial scanning in one radial position have produced experimental values of slip ratio, power (from a travelling incore probe), voidage and coolant mean density over the core height for this position. This one set of distributions has enabled us to test current UKAEA models of subcooled boiling and slip ratio against experiment. From the comparisons, it appears that we can predict the onset of voiding well, but the assumption that a constant fraction of the heat flux forms steam in the subcooled region needs modifying. Of four slip options tested, the current one used by HAMBO and JOSHUA III (Bankoff-Jones) predicts too high a slip ratio. A closer fit to experiment comes from the new Bryce flow-dependent slip option. Any changes in the modelling must be checked, however, with coupled thermal hydraulics-neutronics computations. (author)

  9. Digital control application for the advanced boiling water reactor

    International Nuclear Information System (INIS)

    Fennern, L.E.; Pearson, T.; Wills, H.D.; Swire Rhodes, L.; Pearson, R.L.

    1986-01-01

    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

  10. The interpretation of neutron noise in boiling water reactors

    International Nuclear Information System (INIS)

    John, T.M.; Singh, O.P.

    1985-01-01

    Some qualitative results of neutron noise in a boiling water reactor (BWR) are reported. By using one-group theory, it has been shown that the neutron flux fluctuations caused by a distributed source in space, representative of the coolant boiling noise in BWRs, can be considered as made up of two components: The first one, having a global character, is a quickly varying function of frequency and follows the fundamental mode solution in space; the second, called nonglobal (local), follows the spatial variation of noise-source intensity distribution and is independent of frequency for ω γΣ, this component decreases with increasing frequency. The formulation indicates that the global component is quite sensitive to the neutron multiplication factor of the system and, for the local component, the medium behaves like a nonmultiplying one. The global effect is dominant at lower frequencies in a critical system, and the local effect is dominant at higher fre quencies

  11. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Bouveret, F.

    2001-01-01

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

  12. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

    2010-01-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  13. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

    2010-07-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  14. 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.

  15. Pool Boiling of Hydrocarbon Mixtures on Water

    Energy Technology Data Exchange (ETDEWEB)

    Boee, R.

    1996-09-01

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

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

    International Nuclear Information System (INIS)

    PetrusTakaki, N.

    2012-01-01

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

  17. Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

    Science.gov (United States)

    Gordon, Barry; Garcia, Susan

    Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

  18. Boiling water reactor fuel bundle

    International Nuclear Information System (INIS)

    Weitzberg, A.

    1986-01-01

    A method is described of compensating, without the use of control rods or burnable poisons for power shaping, for reduced moderation of neutrons in an uppermost section of the active core of a boiling water nuclear reactor containing a plurality of elongated fuel rods vertically oriented therein, the fuel rods having nuclear fuel therein, the fuel rods being cooled by water pressurized such that boiling thereof occurs. The method consists of: replacing all of the nuclear fuel in a portion of only the upper half of first predetermined ones of the fuel rods with a solid moderator material of zirconium hydride so that the fuel and the moderator material are axially distributed in the predetermined ones of the fuel rods in an asymmetrical manner relative to a plane through the axial midpoint of each rod and perpendicular to the axis of the rod; placing the moderator material in the first predetermined ones of the fuel rods in respective sealed internal cladding tubes, which are separate from respective external cladding tubes of the first predetermined ones of the fuel rods, to prevent interaction between the moderator material and the external cladding tube of each of the first predetermined ones of the fuel rods; and wherein the number of the first predetermined ones of the fuel rods is at least thirty, and further comprising the steps of: replacing with the moderator material all of the fuel in the upper quarter of each of the at least thirty rods; and also replacing with the moderator material all of the fuel in the adjacent lower quarter of at least sixteen of the at least thirty rods

  19. Outline of the advanced boiling water reactor (ABWR)

    International Nuclear Information System (INIS)

    Hucik, S.A.; Imaoka, T.; Minematsu, A.; Takashima, Y.

    1986-01-01

    The fundamental design of the Advanced Boiling Water Reactor (ABWR) was completed in December 1985. This design represents the next generation of Boiling Water Reactors (BWR) to be introduced into commercial operation in the 1990s. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technologies and many new improvements based on an extensive accumulation of world-wide experience through design, construction and operation of BWRs. The ABWR development program was initiated in 1978, with subsequent design and test and development programs started in 1981. Most of the development and verification tests of the new features have been completed. The ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. The ABWR objectives were specific improvements such as operating and safety margins, enhanced availability and capacity factor, and reduced occupational exposure while at the same time achieving significant cost reduction in both capital and operating costs. The ABWR is characterized by an improved NSSS including ten internal recirculation pumps, fine motion electric-hydraulic control rod drives, optimized safety and auxiliary systems, advanced control and instrumentation systems, improved turbine-generator with moisture/separator reheater with plant output increased to 1350 MWe, and an integrated reinforced concrete containment vessel and compact Reactor and Turbine Building design. The turbine system also included improvements in the Turbine-Generator, feedwater/heater system, and condensate treatment systems. The radwaste system was also optimized taking advantage of the plant design improvements and advances in radwaste technology. The ABWR is a truly optimal design which utilizes advanced technologies, capabilities, performance improvements, and yet provides an economic advantage. (author)

  20. 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.

  1. Spectral shift rod for the boiling water reactor

    International Nuclear Information System (INIS)

    Yokomizo, O.; Kashiwai, S.; Nishida, K.; Orii, A.; Yamashita, J.; Mochida, T.

    1993-01-01

    A Boiling Water Reactor core concept has been proposed using a new fuel component called spectral shift rod (SSR). The SSR is a new type of water rod in which a water level is formed during core operation. The water level can be controlled by the core recirculation flow rate. By using SSRs, the reactor can be operated with all control rods withdrawn through the operation cycle as well as that a much larger natural uranium saving is possible due to spectral shift operation than in current BWRs. The steady state and transient characteristics of the SSRs have been examined by experiments and analyses to certify the feasibility. In a reference design, a four times larger spectral shift width as for the current BWR has been obtained. (orig.)

  2. Investigation of boiling water reactor stability and limit-cycle amplitude

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.A.; Euler, J.A.

    1991-01-01

    Galerkin's method has been applied to a boiling water reactor (BWR) dynamics model consisting of the point kinetics equations, which describe the neutronics, and a feedback transfer function, which describes the thermal hydraulics. The result is a low-order approximate solution describing BWR behavior during small-amplitude limit-cycle oscillations. The approximate solution has been used to obtain a stability condition, show that the average reactor power must increase during limit-cycle oscillations, and qualitatively determine how changes in transfer function values affect the limit-cycle amplitude. 6 refs., 2 figs., 2 tabs

  3. Radiotoxicity study of a boiling water reactor core design based on a thorium-uranium fuel concept

    International Nuclear Information System (INIS)

    Nunez C, A.; Espinosa P, G.

    2007-01-01

    Full text: The innovative design of a Boiling Water Reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233 U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR. A comparison of the toxicity of the spent fuel showed that toxicity is lower in the thorium cycle than other commercial fuels as UO 2 and MOX (uranium and plutonium) in case of the one-through cycle for LWR. (Author)

  4. 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).

  5. Basic philosophy of the safety design of the Toshiba boiling water reactor

    International Nuclear Information System (INIS)

    Sato, T.

    1992-01-01

    This paper discusses the safety design of the Toshiba Boiling Water Reactor (TOSBWR) which was created ∼8 years ago. The design concept is intermediate between conventional boiling water reactors (BWRs) and the advanced BWR (ABWR). It utilizes internal pumps and fine motion control rod drive, but the emergency core cooling system (ECCS) configuration is different from both conventional BWRs and the ABWR. The plant output is 1350 MW (electric). The design is based on two important philosophies: the positive cost reduction philosophy and the constant risk philosophy

  6. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  7. Boiling water reactor modeling capabilities of MMS-02

    International Nuclear Information System (INIS)

    May, R.S.; Abdollahian, D.A.; Elias, E.; Shak, D.P.

    1987-01-01

    During the development period for the Modular Modeling System (MMS) library modules, the Boiling Water Reactor (BWR) has been the last major component to be addressed. The BWRX module includes models of the reactor core, reactor vessel, and recirculation loop. A pre-release version was made available for utility use in September 1983. Since that time a number of changes have been incorporated in BWRX to (1) improve running time for most transient events of interest, (2) extend its capability to include certain events of interest in reactor safety analysis, and (3) incorporate a variety of improvements to the module interfaces and user input formats. The purposes of this study were to briefly review the module structure and physical models, to point the differences between the MMS-02 BWRX module and the BWRX version previously available in the TESTREV1 library, to provide guidelines for choosing among the various user options, and to present some representative results

  8. Corrosion problems in boiling water reactors and their remedies

    International Nuclear Information System (INIS)

    Rosborg, B.

    1989-01-01

    This article briefly presents current corrosion problems in boiling water reactors and their remedies. The problems are different forms of environmentally assisted cracking, and the remedies are divided into material-, environment-, and stress-related remedies. The list of problems comprises: intergranular stress corrosion cracking (IGSCC) in weld-sensitized stainless steel piping; IGSCC in cold-bent stainless steel piping; irradiation-assisted stress corrosion cracking (IASCC) in stainless alloys; IGSCC in high-strength stainless alloys. A prospective corrosion problem, as judged from literature references, and one which relates to plant life, is corrosion fatigue in pressure vessel steel, since the reactor pressure vessel is the most critical component in the BWR pressure boundary as regards plant safety. (author)

  9. Age-related degradation of boiling water reactor vessel internals

    International Nuclear Information System (INIS)

    Ware, A.G.; Shah, V.N.

    1992-01-01

    Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

  10. Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

  11. Dynamic model for a boiling water reactor

    International Nuclear Information System (INIS)

    Muscettola, M.

    1963-07-01

    A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

  12. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  14. Hydrogen water chemistry for boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Cowan, R.L.; Kass, J.N.; Law, R.J.

    1985-01-01

    Hydrogen Water Chemistry (HWC) is now a practical countermeasure for intergranular stress corrosion cracking (IGSCC) susceptibility of reactor structural materials in Boiling Water Reactors (BWRs). The concept, which involves adding hydrogen to the feedwater to suppress the formation of oxidizing species in the reactor, has been extensively studied in both the laboratory and in several operating plants. The Dresden-2 Unit of Commonwealth Edison Company has completed operation for one full 18-month fuel cycle under HWC conditions. The specifications, procedures, equipment, instrumentation and surveillance programs needed for commercial application of the technology are available now. This paper provides a review of the benefits to be obtained, the side affects, and the special operational considerations needed for commercial implementation of HWC. Technological and management ''Lessons Learned'' from work conducted to date are also described

  15. 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)

  16. Efficient characterization of fuel depletion in boiling water reactor

    International Nuclear Information System (INIS)

    Kim, S.H.

    1980-01-01

    An efficient fuel depletion method for boiling water reactor (BWR) fuel assemblies has been developed for fuel cycle analysis. A computer program HISTORY based on this method was designed to carry out accurate and rapid fuel burnup calculation for the fuel assembly. It has been usefully employed to study the depletion characteristics of the fuel assemblies for the preparation of nodal code input data and the fuel management study. The adequacy and the effectiveness of the assessment of this method used in HISTORY were demonstrated by comparing HISTORY results with more detailed CASMO results. The computing cost of HISTORY typically has been less than one dollar for the fuel assembly-level depletion calculations over the full life of the assembly, in contrast to more than $1000 for CASMO. By combining CASMO and HISTORY, a large number of expensive CASMO calculations can be replaced by inexpensive HISTORY. For the depletion calculations via CASMO/HISTORY, CASMO calculations are required only for the reference conditions and just at the beginning of life for other cases such as changes in void fraction, control rod condition and temperature. The simple and inexpensive HISTORY is sufficienty accurate and fast to be used in conjunction with CASMO for fuel cycle analysis and some BWR design calculations

  17. Critical Power Response to Power Oscillations in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Farawila, Yousef M.; Pruitt, Douglas W.

    2003-01-01

    The response of the critical power ratio to boiling water reactor (BWR) power oscillations is essential to the methods and practice of mitigating the effects of unstable density waves. Previous methods for calculating generic critical power response utilized direct time-domain simulations of unstable reactors. In this paper, advances in understanding the nature of the BWR oscillations and critical power phenomena are combined to develop a new method for calculating the critical power response. As the constraint of the reactor state - being at or slightly beyond the instability threshold - is removed, the new method allows the calculation of sensitivities to different operation and design parameters separately, and thus allows tighter safety margins to be used. The sensitivity to flow rate and the resulting oscillation frequency change are given special attention to evaluate the extension of the oscillation 'detect-and-suppress' methods to internal pump plants where the flow rate at natural circulation and oscillation frequency are much lower than jet pump plants

  18. Boiling of subcooled water in forced convection

    International Nuclear Information System (INIS)

    Ricque, R.; Siboul, R.

    1970-01-01

    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/cm 2 ), 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) [fr

  19. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  20. Operating data monitoring and fatigue evaluation systems and findings for boiling water reactors in Japan

    International Nuclear Information System (INIS)

    Maekawa, O.; Kanazawa, Y.; Takahashi, Y.; Tani, M.

    1995-01-01

    The reactor pressure vessel (RPV) is one of the most critical components of a boiling water reactor (BWR) when utilities think about plant life extension (PLEX). Design stress analysis sometimes reports very high fatigue usage factors for such portions of RPVs as stud bolt, feedwater nozzle and support skirt.In order to evaluate design margin and to eliminate excessive conservatism in this design analysis to pave the way for PLEX, Japanese BWR utilities jointly with BWR manufacturers in Japan established a programme (1) to acquire plant operational data on line for specific parameters used in stress analysis, (2) to evaluate margin in the design using measured plant data best estimate boundary conditions for stress analysis, and (3) to establish a simplified fatigue analysis method for BWR RPV.A plant data acquisition system, named OPEDAS, has been developed and installed in Tokyo Electric Power Company's 1100MWe BWR at its Kashiwazaki-Kariwa nuclear power plant. Best estimate stress analysis using measured in-plant data has been carried out and the results show considerable margin in fatigue usage factor over the design. A simplified fatigue analysis method using in-plant data has been developed with the Green's function, although some limitations have been identified for its use. ((orig.))

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

    International Nuclear Information System (INIS)

    1986-09-01

    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

  2. BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hirotaka, E-mail: kawamuh@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (Japan); Hirano, Hideo [Central Research Institute of Electric Power Industry (Japan); Katsumura, Yousuke [University of Tokyo (Japan); Uchida, Shunsuke [Tohoku University (Japan); Mizuno, Takayuki [Mie University (Japan); Kitajima, Hideaki; Tsuzuki, Yasuo [Japan Nuclear Safety Institute (Japan); Terachi, Takumi [Institute of Nuclear Safety System, Inc. (Japan); Nagase, Makoto; Usui, Naoshi [Hitachi-GE Nuclear Energy, Ltd. (Japan); Takagi, Junichi; Urata, Hidehiro [Toshiba Corporation (Japan); Shoda, Yasuhiko; Nishimura, Takao [Mitsubishi Heavy Industry, Ltd. (Japan)

    2016-12-01

    Highlights: • Framework of BWR/PWR water chemistry Guidelines in Japan are presented. • Guideline necessity, definitions, philosophy and technical background are mentioned. • Some guideline settings for control parameters and recommendations are explaines. • Chemistry strategy is also mentioned. - Abstract: After 40 years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved. To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience. This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

  3. 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)

    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

  4. 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.

  5. Variation of the Effectiveness of Hydrogen Water Chemistry in a Boiling Water Reactor during Startup Operations

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya

    2012-09-01

    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 employed in this technology was designed to operate only at power levels greater than 30% of the rated power or at coolant temperatures of greater than 450 deg. F. This 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, and its effectiveness on suppressing SCC initiation was evaluated. 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 normal water chemistry (NWC) or 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] FW s ranging from 0.0 to 2.0 parts per million (ppm) and for power levels ranging from 2.5% 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

  6. Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

    1996-01-01

    A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

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

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Haden, T.H.J.J. van der; Zboray, R.; Manera, A.; Mudde, R.F.

    2002-01-01

    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)

  8. Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Dixon, D.; Lord, D.

    1978-01-01

    The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-11

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

  10. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  11. Leidenfrost boiling of water droplet

    Directory of Open Access Journals (Sweden)

    Orzechowski Tadeusz

    2017-01-01

    Full Text Available The investigations concerned a large water droplet at the heating surface temperature above the Leidenfrost point. The heating cylinder was the main component of experimental stand on which investigations were performed. The measurement system was placed on the high-sensitivity scales. Data transmission was performed through RS232 interface. The author-designed program, with extended functions to control the system, was applied. The present paper examines the behaviour of a large single drop levitating over a hot surface, unsteady mass of the drop, and heat transfer. In computations, the dependence, available in the literature, for the orthogonal droplet projection on the heating surface as a function of time was employed. It was confirmed that the local value of the heat transfer coefficient is a power function of the area of the droplet surface projection. Also, a linear relationship between the flux of mass evaporated from the droplet and the droplet orthogonal projection was observed.

  12. Leidenfrost boiling of water droplet

    Science.gov (United States)

    Orzechowski, Tadeusz

    The investigations concerned a large water droplet at the heating surface temperature above the Leidenfrost point. The heating cylinder was the main component of experimental stand on which investigations were performed. The measurement system was placed on the high-sensitivity scales. Data transmission was performed through RS232 interface. The author-designed program, with extended functions to control the system, was applied. The present paper examines the behaviour of a large single drop levitating over a hot surface, unsteady mass of the drop, and heat transfer. In computations, the dependence, available in the literature, for the orthogonal droplet projection on the heating surface as a function of time was employed. It was confirmed that the local value of the heat transfer coefficient is a power function of the area of the droplet surface projection. Also, a linear relationship between the flux of mass evaporated from the droplet and the droplet orthogonal projection was observed.

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

    International Nuclear Information System (INIS)

    Cowan, R.L.; Head, R.A.; Indig, M.E.; Ruiz, C.P.; Simpson, J.L.

    1988-01-01

    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.)

  14. 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)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  16. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  17. Boiling water reactor containment modeling and analysis at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Holcomb, E.E. III; Wilson, G.E.

    1984-01-01

    Under the auspices of the United States Nuclear Regulatory Commission, severe accidents are being studied at the Idaho National Engineering Laboratory. The boiling water reactor (BWR) studies have focused on postulated anticipated transients without scram (ATWS) accidents which might contribute to severe core damage or containment failure. A summary of the containment studies is presented in the context of the analytical tools (codes) used, typical transient simulation results and the need for prototypical containment data. All of these are related to current and future analytical capabilities. It is shown that torus temperatures during the ATWS depart from limiting conditions for BWR T-quencher operation, outside of which stable steam condensation has not been proven

  18. 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)

    Zisson, P.S.; Whitaker, J.M.; Neilson, H.L.; Mayne, L.L.

    1995-01-01

    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

  19. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

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

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der.

    1989-01-01

    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

  1. Knowledge-Based operation planning system for boiling water reactors

    International Nuclear Information System (INIS)

    Tatsuya Iwamoto; Shungo Sakurai; Hitoshi Uematsu; Makoto Tsuiki

    1987-01-01

    A knowledge-Based Boiling Water Reactor operation planning system was developed to support core operators or core management engineers in making core operation plans, by automatically generating suboptimum core operation procedures. The procedures are obtained by searching a branching tree of the possible core status (nodes) and the elementary operations to change the core status (branches). A path that ends at the target node, and contains only operationally feasible nodes can be a candidate of the solution. The core eigenvalue, the power distribution and the thermal limit parameters at key points are calculated by running a three-dimensional (3-D) BWR core physics simulator to examine the feasibility of the nodes and the performance of candidates. To obtain a practically acceptable solution within a reasonable time rather than making a time-consuming effort to get the optimum one, the Depth-First-Search method, together with the heuristic branch-bounding, was used to search the branching tree. The system was applied to actual operation plannings with real plant data, and gave satisfactory results. It can be concluded that the system can be applied to generate core operation procedures as a substitute for core management experts

  2. 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.

  3. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

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

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

    International Nuclear Information System (INIS)

    Rice, R.E.

    1976-09-01

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

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

    International Nuclear Information System (INIS)

    Garrity, T.F.; Wilkins, D.R.

    1993-01-01

    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

  7. Materials behavior in alternate (hydrogen) water chemistry in the Ringhals-1 boiling water reactor

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Cubicciotti, D.; Trolle, M.

    1986-01-01

    In-plant studies on the intergranular stress corrosion cracking (IGSCC) of sensitized austenitic stainless steel (SS) have been performed at the Swedish Ringhals-1 boiling water reactor (BWR). The studies have covered the present [full-temperature (normal)] water chemistry (PWC) and the alternate (primary) water chemistry (AWC) with hydrogen addition. The test techniques applied were constant extension rate testing (CERT) and electrochemical potential (ECP) measurements. The program was covered by extensive environment monitoring. The results verify earlier laboratory studies which show that sensitized austenitic SS is susceptible to IGSCC in PWC, but not in AWC. Other pressure-bearing BWR construction materials are not adversely affected by AWC. The boundary conditions in Ringhals-1 have been established for an AWC, which is defined as an environment that does not produce IGSCC in sensitized SS. The results are compared with a similar program at Dresden-2, and the points of agreement and discordance in the results are discussed. The relevance of ECP measurements for the control of AWC is discussed

  8. Predicted effect of power uprating on the water chemistry of commercial boiling water reactors

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Chu, Charles F.; Chang Ching

    2009-01-01

    The approach of power uprating has been adopted by operators of light water reactors in the past few decades in order to increase the power generation efficiency of nuclear reactors. The power uprate strategy is apparently applicable to the three nuclear reactors in Taiwan as well. When choosing among the three types of power uprating, measurement uncertainty, stretch power uprating, and extended power uprating, a deliberate and thorough evaluation is required before a final decision and an optimal selection can be made. One practical way of increasing the reactor power is to deliberately adjust the fuel loading pattern and the control rod pattern and thus to avoid replacing the primary coolant pump with a new one of larger capacity. The power density of the reactor will increase with increasing power, but the mass flow rate in the primary coolant circuit (PCC) of a light water reactor will slightly increase (usually by less than 5 %) or even remain unchanged. Accordingly, an uprated power would induce higher neutron and gamma photon dose rates in the reactor coolant but have a minor or no effect on the mass flow rate of the primary coolant. The radiolysis product concentrations and the electrochemical corrosion potential (ECP) values differ largely in the PCC of a boiling water reactor (BWR). It is very difficult to measure the water chemistry data directly at various locations of an actual reactor. Thus the impact of power uprating on the water chemistry of a BWR operating under hydrogen water chemistry (HWC) can only be theoretically evaluated through computer modelling. In this study, the DEMACE computer code was modified to investigate the impact of power uprating on the water chemistry under a fixed mass flow rate in the primary coolant circuit of a BWR/6 type plant. Simulations were carried out for hydrogen concentrations in feedwater ranging from 0.0 to 2.0 mg . kg -1 and for power levels ranging from 100 % to 120 %. The responses of water chemistry and ECP

  9. Hydrogen peroxide and radiation water chemistry of boiling water reactors

    International Nuclear Information System (INIS)

    Ibe, E.; Watanabe, A.; Endo, M.; Takahashi, M.; Karasawa, H.

    1991-01-01

    G-values and rate constants at elevated temperature are reviewed and updated for computer simulation of water radiolysis in BWRs. Quantitative relationship between g-values of H 2 and OH was found out to govern numerically the radiolytic environment in the BWR primary system. Thermal decomposition of hydrogen peroxide was measured in stagnant water in a quartz cell and the rate constant was determined at 2.4 x 10 -7 s -1 with the activation energy of 53.3 kJ/mol. Behaviors of hydrogen peroxide under HWC simulated with updated variables were consistent with plant observation at Forsmark 1 and 2. The most likely decomposition scheme of hydrogen peroxide at surface was identified as H 2 O 2 → H + HO 2 . Based on the surface decomposition process, actual level of hydrogen peroxide was estimated at 200-400 ppb under NWC condition from measured at BWR sampling stations. The estimation was consistent with the numerical simulation of BWR water radiolysis with updated variables. (author)

  10. Influence of subcooled boiling on out-of-phase oscillations in boiling water reactors

    International Nuclear Information System (INIS)

    Munoz-Cobo, J.L.; Chiva, S.; Escriva, A.

    2005-01-01

    In this paper, we develop a reduced order model with modal kinetics for the study of the dynamic behavior of boiling water reactors. This model includes the subcooled boiling in the lower part of the reactor channels. New additional equations have been obtained for the following dynamics magnitudes: the effective inception length for subcooled boiling, the average void fraction in the subcooled boiling region, the average void fraction in the bulk-boiling region, the mass fluxes at the boiling boundary and the channel exit, respectively, and so on. Each channel has three nodes, one of liquid, one with subcooled boiling, and one with bulk boiling. The reduced order model includes also a modal kinetics with the fundamental mode and the first subcritical one, and two channels representing both halves of the reactor core. Also, in this paper, we perform a detailed study of the way to calculate the feedback reactivity parameters. The model displays out-of-phase oscillations when enough feedback gain is provided. The feedback gain that is necessary to self-sustain these oscillations is approximately one-half the gain that is needed when the subcooled boiling node is not included

  11. Variation of the effectiveness of hydrogen water chemistry in a boiling water reactor during power coastdown operations

    International Nuclear Information System (INIS)

    Yeh Tsungkuang; Wang Meiya; Chu, Charles F.; Chang Ching

    2009-01-01

    A theoretical model was adapted to evaluate the impact of power coastdown on the water chemistry of a commercial boiling water reactor (BWR) 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 a commercial BWR 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 a domestic reactor operating under either normal water chemistry or HWC. Our analyses indicated that under a rated core flow rate the chemical 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 level of 90% for Reactor X. (author)

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

    International Nuclear Information System (INIS)

    Nakath, Richard

    2014-01-01

    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.

  13. A boiling-water reactor concept for low radiation exposure based on operating experience

    International Nuclear Information System (INIS)

    Koine, Y.; Uchida, S.; Izumiya, M.; Miki, M.

    1983-01-01

    A review of boiling-water reactor (BWR) operating experience indicates the significant role of water chemistry in determining the radiation dose rate contributing to occupational exposure. The major contributor among the radioactive species involved is identified as 60 Co, produced by neutron activation of 59 Co originating from structural materials. Iron crud, a fine solid form of corrosion product in the reactor water, is also shown to enhance the radiation dose rate. A theoretical study, supported by the operating experience and an extensive confirmatory test, led to the computerized analytical model called DR CRUD which is capable of predicting long-term radiation dose buildup. It accounts for the mechanism of radiation buildup through corrosion products such as irons, cobalts and other radioactive elements; their generation, transport, activation, interaction and deposition in the reactor coolant system are simulated. A scoping analysis, using this model as a tool, establishes the base line of the BWR concept for low occupational exposure. The base line consists of a set of target values for an annual exposure of 200 man.rem in an 1100 MW(e) BWR unit. They are the parameters that will be built into the design such as iron and cobalt inputs to the reactor water, and the capability of the reactor and the condensate purification system. Applicable means of technology are identified to meet the targets, ranging from improved water chemistry to the purification technique, optimized material selection and the recommended operational procedure. Extensive test programmes provide specifications of these means for use in BWRs. Combinations of their application are reviewed to define the concept of reduced exposure. Analytical study verifies the effectiveness of the proposed BWR concept in achieving a low radiation dose rate; occupational exposure is reduced to 200 man.rem/a. (author)

  14. Analysis of a main steam isolation value closure anticipated transient without scram in a boiling water reactor

    International Nuclear Information System (INIS)

    Liaw, T.J.; Pan, C.; Chen, G.S.

    1989-01-01

    Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main steam isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4. Without boron injection and makeup coolant, the reactor loses its coolant inventory very quickly and the reactor power drops rapidly to ∼ 16% of rated power due to negative void reactivity. With coolant makeup from the high-pressure core spray and the reactor core isolation cooling systems, the rector reaches a quasi-steady-state condition after an initially rapidly changing transient. The dome pressure, downcomer water level, and core power oscillate around a mean value; the average core power is ∼ 15%, which is approximately equal to the power needed to heat and evaporate the subcooled makeup coolant. Lower boron concentrations in the core tend to complicate reactor behavior due to the combination of two competing phenomena: the negative boron reactivity and the positive reactivity caused by a void collapse

  15. 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...

  16. PWR and BWR light water reactor systems in the USA and their fuel cycle

    International Nuclear Information System (INIS)

    Crawford, W.D.

    1977-01-01

    Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

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

    International Nuclear Information System (INIS)

    1995-08-01

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

  18. 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.

  19. Development of NUFREQ-N, an analytical model for the stability analysis of nuclear coupled density-wave oscillations in boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Park, G.C.

    1983-01-01

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

  20. Parametric tests of the effects of water chemistry impurities on corrosion of Zr-alloys under simulated BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, S; Ito, K [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Lin, C C [GE Nucklear Energy (United States); Cheng, B [Electric Power Research Inst. (United States); Ikeda, T [Toshiba Corp. (Japan); Oguma, M [Hitachi, Ltd (Japan); Takei, T [Tokyo Electric Power Co., Inc. (Japan); Vitanza, C; Karlsen, T M [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-02-01

    The Halden BWR corrosion test loop was constructed to evaluate the impact of water chemistry variables, heat flux and boiling condition on corrosion performance of Zr-alloys in a simulated BWR environment. The loop consists of two in-core rigs, one for testing fuel rod segments and the other for evaluating water chemistry variables utilizing four miniautoclaves. Ten coupon specimens are enclosed in each miniautoclave. The Zr-alloys for the test include Zircaloy-2 having different nodular corrosion resistance and five new alloys. The first and second of the six irradiation tests planned in this program were completed. Post-irradiation examination of those test specimens have shown that the test loop is capable of producing nodular corrosion on the fuel rod cladding tested under the reference chemistry condition. The miniautoclave tests showed that nodular corrosion could be formed without flux and boiling under some water chemistry conditions and the new alloys, generally, had higher corrosion resistance than the Zircaloy in high oxygen environments. (author). 5 refs, 4 figs, 5 tabs.

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

    International Nuclear Information System (INIS)

    Collins, E.K.; Lai, W.

    1977-01-01

    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

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

    International Nuclear Information System (INIS)

    Karlberg, G.; Goddard, C.; Fitzpatrick, S.

    1994-02-01

    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

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

    International Nuclear Information System (INIS)

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2002-01-01

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

  4. Artificial neural networks versus conventional methods for boiling water reactor stability monitoring

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der

    1995-01-01

    The application of an artificial neural network (ANN) for boiling water reactor (BWR) stability monitoring was studied. A three-layer perceptron was trained on synthetic autocorrelation functions to estimate the decay ratio and the resonance frequency from measured neutron noise. Training of the ANN was improved by adding noise to the training patterns and by applying nonconventional error definitions in the generalized delta rule. The performance of the developed ANN was compared with those of conventional stability monitoring techniques. Explicit care was taken for generating unbiased test data. It is found that the trained ANN is capable of monitoring the stability of the Dodewaard BWR for four specific cases. By comparing properties such as the false alarm ratio, the alarm failure ratio, and the average time to alarm, it is shown that it performs worse than model-based methods in stability monitoring of exact second-order systems but that it is more robust (better resistant to corruptions of the input data and to deviations of the system at issue from an exact second-order system) than other methods. The latter explains its good performance on the Dodewaard BWR and is promising for the application of an ANN for stability monitoring of other reactors and for other operating conditions

  5. Automatic boiling water reactor control rod pattern design using particle swarm optimization algorithm and local search

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-02-15

    Highlights: ► The PSO algorithm was adopted to automatically design a BWR CRP. ► The local search procedure was added to improve the result of PSO algorithm. ► The results show that the obtained CRP is the same good as that in the previous work. -- Abstract: This study developed a method for the automatic design of a boiling water reactor (BWR) control rod pattern (CRP) using the particle swarm optimization (PSO) algorithm. The PSO algorithm is more random compared to the rank-based ant system (RAS) that was used to solve the same BWR CRP design problem in the previous work. In addition, the local search procedure was used to make improvements after PSO, by adding the single control rod (CR) effect. The design goal was to obtain the CRP so that the thermal limits and shutdown margin would satisfy the design requirement and the cycle length, which is implicitly controlled by the axial power distribution, would be acceptable. The results showed that the same acceptable CRP found in the previous work could be obtained.

  6. SBWR: A simplified boiling water reactor

    International Nuclear Information System (INIS)

    Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

    1987-01-01

    An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

  7. 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)

  8. Models and Stability Analysis of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Dorning, John

    2002-01-01

    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

  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. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  11. Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Ibe, Eishi; Nakata, Kiyatomo; Fuse, Motomasa; Ohsumi, Katsumi; Takashima, Yoshie

    1995-01-01

    Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the crack growth rate, even if microcracks are present on the structural materials. Several disadvantage caused by HWC are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated base on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of Hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants -- e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom -- are planned that would collect data to support the theoretical considerations

  12. Stability monitoring of a boiling water reactor

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der; Hoogenboom, J.E.

    1989-01-01

    Stability monitoring is of great importance for optimal plant performance. Decay ratios for several operating conditions show that the Dodewaard BWR is very stable and that pressure lowering, power increase and flux peaking lead to a higher decay ratio (worse stability). 1 fig., 1 tab., 1 ref

  13. 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.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  15. 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)

    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

  16. 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.

  17. A non-linear reduced order methodology applicable to boiling water reactor stability analysis

    International Nuclear Information System (INIS)

    Prill, Dennis Paul

    2013-01-01

    Thermal-hydraulic coupling between power, flow rate and density, intensified by neutronics feedback are the main drivers of boiling water reactor (BWR) stability behavior. High-power low-flow conditions in connection with unfavorable power distributions can lead the BWR system into unstable regions where power oscillations can be triggered. This important threat to operational safety requires careful analysis for proper understanding. Analyzing an exhaustive parameter space of the non-linear BWR system becomes feasible with methodologies based on reduced order models (ROMs), saving computational cost and improving the physical understanding. Presently within reactor dynamics, no general and automatic prediction of high-dimensional ROMs based on detailed BWR models are available. In this thesis a systematic self-contained model order reduction (MOR) technique is derived which is applicable for several classes of dynamical problems, and in particular to BWRs of any degree of details. Expert knowledge can be given by operational, experimental or numerical transient data and is transfered into an optimal basis function representation. The methodology is mostly automated and provides the framework for the reduction of various different systems of any level of complexity. Only little effort is necessary to attain a reduced version within this self-written code which is based on coupling of sophisticated commercial software. The methodology reduces a complex system in a grid-free manner to a small system able to capture even non-linear dynamics. It is based on an optimal choice of basis functions given by the so-called proper orthogonal decomposition (POD). Required steps to achieve reliable and numerical stable ROM are given by a distinct calibration road-map. In validation and verification steps, a wide spectrum of representative test examples is systematically studied regarding a later BWR application. The first example is non-linear and has a dispersive character

  18. Two-loop feed water control system in BWR plants

    International Nuclear Information System (INIS)

    Omori, Takashi; Watanabe, Takao; Hirose, Masao.

    1982-01-01

    In the process of the start-up and shutdown of BWR plants, the operation of changing over feed pumps corresponding to plant output is performed. Therefore, it is necessary to develop the automatic changeover system for feed pumps, which minimizes the variation of water level in reactors and is easy to operate. The three-element control system with the water level in reactors, the flow rate of main steam and the flow rate of feed water as the input is mainly applied, but long time is required for the changeover of feed pumps. The two-loop feed control system can control simultaneously two pumps being changed over, therefore it is suitable to the automatic changeover control system for feed pumps. Also it is excellent for the control of the recirculating valves of feed pumps. The control characteristics of the two-loop feed water control system against the external disturbance which causes the variation of water level in reactors were examined. The results of analysis by simulation are reported. The features of the two-loop feed water control system, the method of simulation and the evaluation of the two-loop feed water control system are described. Its connection with a digital feed water recirculation control system is expected. (Kako, I.)

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

    International Nuclear Information System (INIS)

    Gluntz, D.M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs

  20. The mechanisms of transitions from natural convection and nucleate boiling to nucleate boiling or film boiling caused by rapid depressurization in highly subcooled water

    International Nuclear Information System (INIS)

    Sakurai, Akira; Shiotsu, Masahiro; Hata, Koichi; Fukuda, Katsuya

    1999-01-01

    The mechanisms of transient boiling process including the transitions to nucleate boiling or film boiling from initial heat fluxes, q in , in natural convection and nucleate boiling regimes caused by exponentially decreasing system pressure with various decreasing periods, τ p on a horizontal cylinder in a pool of highly subcooled water were clarified. The transient boiling processes with different characteristics were divided into three groups for low and intermediate q in in natural convection regime, and for high q in in nucleate boiling regime. The transitions at maximum heat fluxes from low q in in natural convection regime to stable nucleate boiling regime occurred independently of the τ p values. The transitions from intermediate and high q in values in natural convection and nucleate boiling to stable film boiling occurred for short τ p values, although those to stable nucleate boiling occurred for tong τ p values. The CHF and corresponding surface superheat values at which the transition to film boiling occurred were considerably lower and higher than the steady-state values at the corresponding pressure during the depressurization respectively. It was suggested that the transitions to stable film boiling at transient critical heat fluxes from intermediate q in in natural convection and from high q in in nucleate boiling for short τ p occur due to explosive-like heterogeneous spontaneous nucleation (HSN). The photographs of typical vapor behavior due to the HSN during depressurization from natural convection regime for short τ p were shown. (author)

  1. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2001-06-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current nuclear applications Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for the purpose. The present volume describes the specification of such a benchmark. The transient addressed is a turbine trip (TT) in a BWR involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the plant make the present benchmark very valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. Modelling of crack chemistry in sensitized stainless steel in boiling water reactor environments

    International Nuclear Information System (INIS)

    Turnbull, A.

    1997-01-01

    An advanced model has been used to predict the chemistry and potential in a stress corrosion crack in sensitized stainless steel in a boiling water reactor (BWR) environment. The model assumes trapezoidal crack geometry, incorporates anodic reaction and cathodic reduction within the crack, and takes into account the limited solubility of cations in high temperature water. The results indicate that the crack tip potential is not independent of the external potential, and that the reactions on the walls of the crack must be included for reliable prediction. Accordingly, both the modelling assumptions of Ford and Andresen and of Macdonald and Urquidi-Macdonald, whilst having merit, are not fully satisfactory. Extended application of the model for improved prediction of stress corrosion crack growth rate is constrained by limitations in electrochemical data which are currently inadequate. (author)

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

    International Nuclear Information System (INIS)

    Smith, P.F.

    1992-01-01

    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. Study on water boiling noises in a large volume

    International Nuclear Information System (INIS)

    Masagutov, R.F.; Krivtsov, V.A.

    1977-01-01

    Presented are the results of measurement of the noise spectra during boiling of water in a large volume at the pressure of 1 at. Boiling of the distilled water has been accomplished with the use of the heaters made of the Kh18N10T steel, 50 mm in length, 2 mm in the outside diameter, with the wall thickness of 0.1 mm. The degree of water under heating changed during the experiments from 0 to 80 deg C, and the magnitude of the specific heat flux varied from o to 0.7 - 0.9 qsup(x), where qsup(x) was the specific heat flux of the tube burn-out. The noise spectrum of the boiling water was analyzed at frequencies of 0.5 to 200 kHz. The submerge-type pressure-electric transmitters were used for measurements. At underheating boiling during the experiment the standing waves have formed which determine the structure of the measured spectra. During saturated boiling of water no standing waves were revealed. At underheating over 15 - 20 deg C the water boiling process is accompanied by the noises within the ultrasonic frequency range. The maximum upper boundary of the noise in the experiments amounts to 90 - 100 kHz

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

    International Nuclear Information System (INIS)

    Nunez-Carrera, Alejandro; Francois, Juan Luis; Martin-del-Campo, Cecilia; Espinosa-Paredes, Gilberto

    2005-01-01

    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 233 U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235 U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly

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

    International Nuclear Information System (INIS)

    Stellwag, B.; Laendner, A.; Weiss, S.; Huettner, F.

    2010-01-01

    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)

  8. 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.

  9. Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

    2017-05-15

    Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

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

    International Nuclear Information System (INIS)

    1994-01-01

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

    International Nuclear Information System (INIS)

    Diaz S, A.; Robles, E.F.

    2003-01-01

    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. A symptom based decision tree approach to boiling water reactor emergency operating procedures

    International Nuclear Information System (INIS)

    Knobel, R.C.

    1984-01-01

    This paper describes a Decision Tree approach to development of BWR Emergency Operating Procedures for use by operators during emergencies. This approach utilizes the symptom based Emergency Procedure Guidelines approved for implementation by the USNRC. Included in the paper is a discussion of the relative merits of the event based Emergency Operating Procedures currently in use at USBWR plants. The body of the paper is devoted to a discussion of the Decision Tree Approach to Emergency Operating Procedures soon to be implemented at two United States Boiling Water Reactor plants, why this approach solves many of the problems with procedures indentified in the post accident reviews of Three Mile Island procedures, and why only now is this approach both desirable and feasible. The paper discusses how nuclear plant simulators were involved in the development of the Emergency Operating Procedure decision trees, and in the verification and validation of these procedures. (orig./HP)

  13. Mark I 1/5-scale boiling water reactor pressure suppression experiment. Quick-look report for test numbers 1.0(a) and 1.0(b) performed on March 4 and 8, 1977

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  14. BWR internals life assurance

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  15. Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

    International Nuclear Information System (INIS)

    Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

    2004-01-01

    Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

  16. Void fraction and incipient point of boiling during the subcooled nucleate flow boiling of water

    International Nuclear Information System (INIS)

    Unal, H.C.

    1977-01-01

    Void fraction has been determined with high-speed photography for subcooled nucleate flow boiling of water. The data obtained and the data of various investigators for adiabatic flow of stream-water mixtures and saturated bulk boiling of water have yielded a correlation which covers the following conditions: geometry: vertically orientated circular tubes, rectangular channels and annuli; pressure: 2 to 15.9 MN/m 2 ; mass velocity: 388 to 3500 kg/m 2 s; void fraction: 0 to 99%; hydraulic diameter: 0.0047 to 0.0343 m; heat flux: adiabatic and 0.01 to 2.0 MW/m 2 . The accuracy of the correlation is estimated to be 12.5%. The value of the so-called distribution (or flow) parameter has been experimentally determined and found to be equal to 1 for a vertical small-diameter circular tube. The incipient point of boiling for subcooled nucleate flow boiling of water has been determined with high-speed photography. The data obtained and the data available in the literature have yielded a correlation which covers the following conditions: geometry: plate, circular tube and inner tube-heated, outer tube-heated and inner - and outer tube heated annulus; pressure: 0.15 to 15.9 MN/m 2 ; mass velocity: 470 to 17355 kg/m 2 s; hydraulic diameter: 0.00239 to 0.032 m; heat flux: 0.13 to 9.8 MW/m 2 ; subcooling: 2.6 to 108 K; material of heating surface: stainless steel and nickel. The accuracy of the correlation is estimated to be 27.5%. Maximum bubble diameters have been measured at the incipient point of boiling. These data and the data from literature have been correlated for the pressure range of 0.1 to 15.9 MN/m 2 . (author)

  17. Flow film boiling heat transfer in water and Freon-113

    International Nuclear Information System (INIS)

    Liu, Qiusheng; Shiotsu, Masahiro; Sakurai, Akira

    2002-01-01

    Experimental apparatus and method for film boiling heat transfer measurement on a horizontal cylinder in forced flow of water and Freon-113 under pressurized and subcooled conditions were developed. The experiments of film boiling heat transfer from single horizontal cylinders with diameters ranging from 0.7 to 5 mm in saturated and subcooled water and Freon-113 flowing upward perpendicular to the cylinders were carried out for the flow velocities ranging from 0 to 1 m/s under system pressures ranging from 100 to 500 kPa. Liquid subcoolings ranged from 0 to 50 K, and the cylinder surface superheats were raised up to 800 K for water and 400 K for Freon-113. The film boiling heat transfer coefficients obtained were depended on surface superheats, flow velocities, liquid subcoolings, system pressures and cylinder diameters. The effects of these parameters were systematically investigated under wider ranges of experimental conditions. It was found that the heat transfer coefficients are higher for higher flow velocities, subcoolings, system pressures, and for smaller cylinder diameters. The observation results of film boiling phenomena were obtained by a high-speed video camera. A new correlation for subcooled flow film boiling heat transfer was derived by modifying authors' correlation for saturated flow film boiling heat transfer with authors' experimental data under wide subcooled conditions. (author)

  18. Drift flux formulation of a boiling water reactor channel with subcooled boiling

    International Nuclear Information System (INIS)

    Elias, E.E.; Shak, D.P.; May, R.S.

    1987-01-01

    The channel formulation used in the BWR module of the Modular Modeling System MMS-02 is presented. The purpose of channel model is to accurately predict the transient response of the enthalpy void and flow rate. Accurate prediction of the two-phase enthalpy, and void fraction distributions along the channel is important since they are key input parameters to the neutronic model, and have direct effect on the core and overall reactor response. In order to model the channel response correctly, the physical phenomena had to be realistically represented. The model accounts for subcooled boiling and slip through the use of an empirical subcooled void-quality model. Simplifying assumptions are made so that only one differential equation, the energy equation, is integrated along the channel. A consistent use of semi-empirical correlations enables a complete representation of the channel flow and void fraction with the bulk enthalpy as the only state variable. The differential equation and the constitutive relations of this two-phase flow model are presented. Several numerical examples are given, and finally, come conclusions are presented

  19. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

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

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation 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 that are invariant. These groupings 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 and occurs in a time reduced by the square root of the scale factor

  20. Numerical simulation of progressive inlet orifices in boiling water reactor fuel

    International Nuclear Information System (INIS)

    Lundgren, Sara

    2004-07-01

    This thesis was carried out at Forsmark Nuclear Power Plant. The power plant in Forsmark consists of three boiling water reactors (BWR) which produce about 17% of Swedish electricity. In a BWR the nuclear reactions are used to boil water inside the reactor vessel. The water works both as a coolant and as a moderator and the resulting steam is used directly to run the turbines. A problem when running a BWR at low flow conditions is the density wave oscillations that might occur to the water flow inside the fuel assemblies. These oscillations arise due to the connection between power and flow rate in a heated channel with two-phase flow. In order to improve the stability performance of the channel an orifice plate is placed at the inlet of each fuel assembly. Today these orifice plates have sharp edges and a constant resistance coefficient. Experimental work has been done with progressive orifices, the edge of which is half-oval in shape. The advantage of progressive orifices is the lower pressure losses with an increase of the Reynolds number, a similar phenomenon that appears in external flow around curved bodies. Since there are high costs associated with experimental generation of high- temperature and high-pressure data, it is of some interest to be able to reproduce and generate data using Computational Fluid Dynamics (CFD). This work deals with the possibility to use the CFD-code Fluent to do numerical simulations of the flow through progressive orifices. The following conclusions may be drawn from the numerical results: All simulations using Reynolds-Averaged Navier-Stokes (RANS) turbulence models, two-dimensional and three-dimensional, capture an abrupt decrease of the resistance coefficient at higher Reynolds numbers. Two-equation models seem to under-predict the critical Reynolds number. The five-equation Reynolds Stress Model (RSM) gives a critical Reynolds number of the same order of magnitude of that measured in experiments. No major differences have

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

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

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

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

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1991-08-01

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

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

    International Nuclear Information System (INIS)

    Motoda, H.; Tanisaka, S.; Kiguchi, T.; Yonenaga, H.

    1977-01-01

    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

  4. 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.

  5. Higher order generalized perturbation theory for boiling water reactor in-core fuel management optimization

    International Nuclear Information System (INIS)

    Moore, B.R.; Turinsky, P.J.

    1998-01-01

    Boiling water reactor (BWR) loading pattern assessment requires solving the two-group, nodal form of the neutron diffusion equation and drift-flux form of the fluid equations simultaneously because these equation sets are strongly coupled via nonlinear feedback. To reduce the computational burden associated with the calculation of the core attributes (that is, core eigenvalue and thermal margins) of a perturbed BWR loading pattern, the analytical and numerical aspects of a higher order generalized perturbation theory (GPT) method, which correctly addresses the strong nonlinear feedbacks of two-phase flow, have been established. Inclusion of Jacobian information in the definition of the generalized flux adjoints provides for a rapidly convergent iterative method for solution of the power distribution and eigenvalue of a loading pattern perturbed from a reference state. Results show that the computational speedup of GPT compared with conventional forward solution methods demanding consistent accuracy is highly dependent on the number of spatial nodes utilized by the core simulator, varying from superior to inferior performance as the number of nodes increases

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

    International Nuclear Information System (INIS)

    Sonoda, Y.; Kanemoto, S.; Imaruoka, H.

    1990-01-01

    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

  7. 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)

    Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

    2004-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    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)

  9. Study of the radiotoxicity of actinides recycling in boiling water reactors fuel

    International Nuclear Information System (INIS)

    Francois, J.L.; Guzman, J.R.; Martin-del-Campo, C.

    2009-01-01

    In this paper the production and destruction, as well as the radiotoxicity of plutonium and minor actinides (MA) obtained from the multi-recycling of boiling water reactors (BWR) fuel are analyzed. A BWR MOX fuel assembly, with uranium (from enrichment tails), plutonium and minor actinides is designed and studied using the HELIOS code. The actinides mass and the radiotoxicity of the spent fuel are compared with those of the once-through or direct cycle. Other type of fuel assembly is also analyzed: an assembly with enriched uranium and minor actinides; without plutonium. For this study, the fuel remains in the reactor for four cycles, where each cycle is 18 months length, with a discharge burnup of 48 MWd/kg. After this time, the fuel is placed in the spent fuel pool to be cooled during 5 years. Afterwards, the fuel is recycled for the next fuel cycle; 2 years are considered for recycle and fuel fabrication. Two recycles are taken into account in this study. Regarding radiotoxicity, results show that in the period from the spent fuel discharge until 1000 years, the highest reduction in the radiotoxicity related to the direct cycle is obtained with a fuel composed of MA and enriched uranium. However, in the period after few thousands of years, the lowest radiotoxicity is obtained using the fuel with plutonium and MA. The reduction in the radiotoxicity of the spent fuel after one or two recycling in a BWR is however very small for the studied MOX assemblies, reaching a maximum reduction factor of 2.

  10. BWR fuel performance under advanced water chemistry conditions – a delicate journey towards zero fuel failures – a review

    International Nuclear Information System (INIS)

    Hettiarachchi, S.

    2015-01-01

    Boiling Water Reactors (BWRs) have undergone a variety of chemistry evolutions over the past few decades as a result of the need to control stress corrosion cracking of reactor internals, radiation fields and personnel exposure. Some of the advanced chemistry changes include hydrogen addition, zinc addition, iron reduction using better filtration technologies, and more recently noble metal chemical addition to many of the modern day operating BWRs. These water chemistry evolutions have resulted in changes in the crud distribution on fuel cladding material, Co-60 levels and the Rod oxide thickness (ROXI) measurements using the conventional eddy current techniques. A limited number of Post-Irradiation Examinations (PIE) of fuel rods that exhibited elevated oxide thickness using eddy current techniques showed that the actual oxide thickness by metallography is much lower. The difference in these observations is attributed to the changing magnetic properties of the crud affecting the rod oxide thickness measurement by the eddy current technique. This paper will review and summarize the BWR fuel cladding performance under these advanced and improved water chemistry conditions and how these changes have affected the goal to reach zero fuel failures. The paper will also provide a brief summary of some of the results of hot cell PIE, results of crud composition evaluation, crud spallation, oxide thickness measurements, hydrogen content in the cladding and some fuel failure observations. (author) Key Words: Boiling Water Reactor, Fuel Performance, Hydrogen Addition, Zinc Addition, Noble Metal Chemical Addition, Zero Leakers

  11. Operating experience of natural circulation core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    Kullberg, C.; Jones, K.; Heath, C.

    1993-01-01

    General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

  12. ECP measurements in the BWR-1 water loop relative to water composition changes

    Energy Technology Data Exchange (ETDEWEB)

    Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2010-07-01

    The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

  13. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

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

    International Nuclear Information System (INIS)

    Queiser, H.

    1976-01-01

    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.) [de

  15. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Belblidia, L.A.; Chevrier, A.

    1994-01-01

    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

  16. Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)], E-mail: Zoran.Stosic@areva.com; Brettschuh, Werner; Stoll, Uwe [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)

    2008-08-15

    AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of

  17. Electrochemical corrosion potential monitoring in boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Hettiarachchi, S.; Hale, D.H.; Law, R.J.

    1998-01-01

    The electrochemical corrosion potential (ECP) is defined as the measured voltage between a metal and a standard reference electrode converted to the standard hydrogen electrode (SHE) scale. This concept is shown schematically in Figure 1. The measurement of ECP is of primary importance for both evaluating the stress corrosion cracking susceptibility of a component and for assuring that the specification for hydrogen water chemistry, ECP < -230 mV, SHE is being met. In practice, only a limited number of measurement locations are available in the BWR and only a few reference electrode types are robust enough for BWR duty. Because of the radiolysis inherent in the BWR, local environment plays an important role in establishing the ECP of a component. This paper will address the strategies for obtaining representative measurements, given these stated limitations and constraints. The paper will also address the ECP monitoring strategies for the noble metal chemical addition process that is being implemented in BWRs to meet the ECP specification at low hydrogen injection rates. (author)

  18. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

  19. 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)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  1. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

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

  2. Construction of the advanced boiling water reactor in Japan

    International Nuclear Information System (INIS)

    Natsume, Nobuo; Noda, Hiroshi

    1996-01-01

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

  3. Power plant design: ESBWR - the latest passive BWR

    International Nuclear Information System (INIS)

    Arnold, H.; Yadigaroglu, G.; Stoop, P.C.

    1997-01-01

    When General Electric said it would end development of its 670 MWe SBWR (Simplified Boiling Water Reactor), it was not quite the end of the story. Also on the drawing board at the time was the larger ESBWR (standing for either European or Economic Simplified BWR) whose goal was to provide the improved economic performance that the SBWR could not. (UK)

  4. Development of an in-core fuel management tool for boiling water reactors

    International Nuclear Information System (INIS)

    Gilli, Luca; Wakker, Pieter H.; Elder, Brian R.

    2017-01-01

    The in-core fuel management of a nuclear reactor is a challenging task due to the virtually infinite number of loading patterns one could theoretically adopt. The ROSA (Reloading Optimization by Simulated Annealing) code is an optimization tool that has been successfully used in the last two decades to facilitate the core design of several Pressurized Water Reactors (PWRs). It is designed to perform a stochastic search for an optimal Loading Pattern (LP) using a simulated annealing algorithm. This corresponds to performing a depletion calculation for each one of the hundreds of thousands of unique LPs generated during the stochastic search. Therefore, speed is one of the most important requirements that the solvers used by the depletion tool must fulfill. ROSA's depletion analysis tool makes use of a particularly fast nodal method (known as the kernel method) for the evaluation of the power distribution associated with a particular LP. One of the strongest assumptions behind the kernel method is that the neutron migration length does not change considerably between the point where a neutron is generated and the point where the same neutron is absorbed. Although strong, this assumption is quite compatible with the neutronic characteristics of PWRs cores. In this paper we give an overview of the work done in order to develop a version of ROSA capable of performing the core design of Boiling Water Reactors (BWRs). We focus the discussion on the development of the depletion analysis tool by outlining the modifications of the kernel methods implemented in order to make the solver accurate for BWR cores. An improvement of the definition of the transport kernel is necessary to take the strong anisotropies characterizing the neutronic problem into account. These anisotropies arise due to the presence of strong changes in the moderator density and due to the presence of control blades. Furthermore, we are going to discuss how the boundary conditions are adopted by the

  5. Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Conlon, W.M.; Lahey, R.T. Jr.

    1982-12-01

    This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

  6. Noise investigations in boiling water and pressurized water reactors

    International Nuclear Information System (INIS)

    Kosaly, G.

    1979-08-01

    The paper begins with the general discussion of the neutron noise measured by in-core and ex-core detectors. The theory of BWR noise is surveyed in detail. After the theoretical survey the physical background of the measuring methods applied in the fluctuation analysis of BWRs is discussed. The section on PWR noise discusses the information content of the neutron, pressure and displacement signals about the vibration of the vessel and its internals. (author)

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

    International Nuclear Information System (INIS)

    Orlov, A.

    2011-01-01

    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 ZrO 2 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 (M x Fe y )[M (1-x) Fe (2-y) ]O 4 , 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 (ZnFe 2 O 4 , NiFe 2 O 4 and MnFe 2 O 4 ) proving the existence of

  8. Improvement of steam separator in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, Jan Peter; Cremer, Ingo; Lorenz, Maik [AREVA GmbH, Erlangen (Germany)

    2013-07-01

    The potential to improve the function of the steam separator is identified and explored by scaled air-water tests and validated CFD calculations. It can be outlined that requirements on a modern steam separator for BWR plants will be fulfilled, combined with very good operational experience of the existing separator designs (e.g. material, layout). With the new steam separator design, modern high performance fuel assembly designs can be used for various core loading strategies (e.g. low leakage). This allows an increased thermal power of up to +50 % for the fuel element clusters in the center of the core with high radial peaking factors. In addition, any problems with unallowable high moisture at the turbine are solved with the new design, which have been identified for running BWR plants with the old steam separator design after applying new core loading patterns (e.g. after power uprates). A compatible steam separator design for all running BWRs is ready to launch. (orig.)

  9. Controllability studies for an advanced CANDU boiling light water reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Hinds, H.W.

    1976-12-01

    Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

  10. Update on materials performance and electrochemistry in hydrogen water chemistry at Dresden-2 BWR

    International Nuclear Information System (INIS)

    Indig, M.E.; Weber, J.E.; Davis, R.B.; Gordon, B.M.

    1985-01-01

    Previous studies performed in 1982 indicated that if sufficient hydrogen was injected into the Dresden-2 BWR, IGSCC of sensitized austenitic stainless steel was mitigated. The present series of experiments were aimed at verification of the above finding, determining how much time off hydrogen water chemistry (HWC) could be tolerated and how HWC affected pre-existing cracks

  11. Development of advanced boiling water reactor for medium capacity

    International Nuclear Information System (INIS)

    Kazuo Hisajima; Yutaka Asanuma

    2005-01-01

    This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

  12. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  13. Instrument lance for boiling water reactors

    International Nuclear Information System (INIS)

    Proell, N.; Bertz, S.; Graebener, K.H.

    1980-01-01

    The instrument lance contains in the lance cover pipe a thimble as part of the drive chamber system. Other thimbles serve to carry neutron detectors. Detectors can be exchanged without opening the reactor pressure vessel and without removing the fuel elements. Furthermore the detector exchange is independent from the fuel element cycle. The measurement lance passes from the bottom of the pressure vessel over the total hight of the core in the water ducts between the fuel elements and can thus determine the neutron flux distribution. (DG) [de

  14. A study of implementing In-Cycle-Shuffle strategy to a decommissioning boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chung-Yuan, E-mail: tuckjason@iner.gov.tw; Tung, Wu-Hsiung; Yaur, Shyun-Jung

    2017-06-15

    Highlights: • A loading pattern strategy ICS (In-Cycle-Shuffle) was implemented to the last cycle of the boiling water reactor. • The best power sharing distribution and ICS timing was found. • A new parameter “Burnup sharing” is presented to evaluate ICS strategy. - Abstract: In this paper, a loading pattern strategy In-Cycle-Shuffle (ICS) is implemented to the last cycle of the boiling water reactor (BWR) before decommissioning to save the fuel cycle cost. This method needs a core shutdown during the operation of a cycle to change the loading pattern to gain more reactivity. The reactivity model is used to model the ICS strategy in order to find out the best ICS timing and the optimum power sharing distribution before ICS and after ICS. Several parameters of reactivity model are modified and the effect of burnable poison, gadolinium (Gd), is considered in this research. Three cases are presented and it is found that the best ICS timing is at about two-thirds of total cycle length no matter the poisoning effect of Gd is considered or not. According to the optimum power sharing distribution result, it is suggested to decrease the once burnt power and increase the thrice burnt fuel power as much as possible before ICS. After ICS, it is suggested to increase the positive reactivity fuel power and decrease the thrice burnt fuel power as much as possible. A new parameter “Burnup sharing” is presented to evaluate the special case whose EOC power weighting factor and the burnup accumulation factor in the reactivity model are quite different.

  15. A study of implementing In-Cycle-Shuffle strategy to a decommissioning boiling water reactor

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shyun-Jung

    2017-01-01

    Highlights: • A loading pattern strategy ICS (In-Cycle-Shuffle) was implemented to the last cycle of the boiling water reactor. • The best power sharing distribution and ICS timing was found. • A new parameter “Burnup sharing” is presented to evaluate ICS strategy. - Abstract: In this paper, a loading pattern strategy In-Cycle-Shuffle (ICS) is implemented to the last cycle of the boiling water reactor (BWR) before decommissioning to save the fuel cycle cost. This method needs a core shutdown during the operation of a cycle to change the loading pattern to gain more reactivity. The reactivity model is used to model the ICS strategy in order to find out the best ICS timing and the optimum power sharing distribution before ICS and after ICS. Several parameters of reactivity model are modified and the effect of burnable poison, gadolinium (Gd), is considered in this research. Three cases are presented and it is found that the best ICS timing is at about two-thirds of total cycle length no matter the poisoning effect of Gd is considered or not. According to the optimum power sharing distribution result, it is suggested to decrease the once burnt power and increase the thrice burnt fuel power as much as possible before ICS. After ICS, it is suggested to increase the positive reactivity fuel power and decrease the thrice burnt fuel power as much as possible. A new parameter “Burnup sharing” is presented to evaluate the special case whose EOC power weighting factor and the burnup accumulation factor in the reactivity model are quite different.

  16. Water chemistry and radiation buildup at the Commonwealth Edison Company LaSalle-1 BWR. Final report

    International Nuclear Information System (INIS)

    Earls, C.E.; Blok, J.

    1986-09-01

    This report presents the results of a comprehensive study of the water quality and radiation buildup at the LaSalle County Unit 2 boiling warer reactor (BWR). The purpose of the study was to determine the effect of corrosion product inputs from the forward pumped heater drains on overall water quality. Since the drains are pumped into the feedwater line without filtration or demineralization, corrosion products in these streams will directly add to the impurity levels of the final feedwater. At LaSalle, the forward pumped heater drains contributed less to the feedwater impurities, on average, than the effluent of the condensate demineralizer. The feedwater quality at LaSalle was generally in the ''acceptable'' range. Nevertheless, significant water chemistry improvements, especially in reducing the corrosion product spikes associated with power or flow transients, is highly desirable for this plant. Such improvements should begin with a more consistent quality of demineralizer operation. Quantitative gamma scans of the primary system piping at LaSalle 2 were carried out in the course of the water chemistry study. Although the cumulative operational exposure of the plant was relatively limited at the time this study was carried out, the radiation buildup rate did appear to be rapid (in fact, among the most rapid) compared to other similar BWRs

  17. GPE-BWR and the containment venting and filtering issue

    International Nuclear Information System (INIS)

    Palomo, J.; Santiago, J. de

    1988-01-01

    The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

  18. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  19. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

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

    International Nuclear Information System (INIS)

    Chabert, Christine

    1994-01-01

    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 Pu 239 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 U 235 , the production of plutonium and burnup, and a better estimation of the material balance. (author) [fr

  1. A simplified, coarse-mesh, three-dimensional diffusion scheme for calculating the gross power distribution in a boiling water reactor

    International Nuclear Information System (INIS)

    Borresen, S.

    1995-01-01

    A simplified, finite-difference diffusion scheme for a three-dimensional calculation of the gross power distribution in the core of a boiling water reactor (BWR) is presented. Results obtained in a series of one- and two-dimensional test cases indicate that this method may be of sufficient accuracy and simplicity for implementation in BWR-simulator computer programs. Computer requirements are very modest; thus, only 3N memory locations are required for in-core treatment of the inner iteration in the solution of a problem with N mesh points. The mesh width may be chosen equal to the fuel assembly pitch. Input data are in the form of conventional 2-group diffusion parameters. It is concluded that the method presented has definite advantages in comparison with the nodal coupling method. (author)

  2. 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.

  3. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    Science.gov (United States)

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  4. Predictors of Drinking Water Boiling and Bottled Water Consumption in Rural China: A Hierarchical Modeling Approach.

    Science.gov (United States)

    Cohen, Alasdair; Zhang, Qi; Luo, Qing; Tao, Yong; Colford, John M; Ray, Isha

    2017-06-20

    Approximately two billion people drink unsafe water. Boiling is the most commonly used household water treatment (HWT) method globally and in China. HWT can make water safer, but sustained adoption is rare and bottled water consumption is growing. To successfully promote HWT, an understanding of associated socioeconomic factors is critical. We collected survey data and water samples from 450 rural households in Guangxi Province, China. Covariates were grouped into blocks to hierarchically construct modified Poisson models and estimate risk ratios (RR) associated with boiling methods, bottled water, and untreated water. Female-headed households were most likely to boil (RR = 1.36, p water, or use electric kettles if they boiled. Our findings show that boiling is not an undifferentiated practice, but one with different methods of varying effectiveness, environmental impact, and adoption across socioeconomic strata. Our results can inform programs to promote safer and more efficient boiling using electric kettles, and suggest that if rural China's economy continues to grow then bottled water use will increase.

  5. Verification of the CASMO-3/SIMULATE-3 pin power accuracy by comparison with operating boiling water reactor measurements

    International Nuclear Information System (INIS)

    Uegata, T.; Saji, E.; Tanaka, H.

    1993-01-01

    Intranodal pin power distributions calculated by the CASMO-3/SIMULATE-3 code have been compared with pin gamma scan measurements. These data were obtained from the depleted core of an operating boiling water reactor (BWR), which is more complicated than a pressurized water reactor to calculate because of the existence of coolant void distributions and cruciform control blades. Furthermore, measured bundles include mixed-oxide (MOX) bundles in which steep thermal flux gradients occur. Both UO 2 and MOX bundles have been calculated in the same manner based on the standard CASMO-3/SIMULATE-3 methods. The total pin power root-mean-square (rms) error is 2.7%, which includes measurement error, from an 896-point comparison. There is no obvious dependency on axial elevations (void fractions) and no significant difference between fuel types (UO 2 or MOX), although the errors in a peripheral bundle, which is less important from the standpoint of core design, are somewhat larger than those in the internal bundles. If the peripheral bundle is excluded, the total rms error is reduced to 2.2%. From these results, it is concluded that excellent agreement has been obtained between the calculations and measurements and that the calculational capability of CASMO-3/SIMULATE-3 for the intranodal pin power distribution is quite satisfactory and useful for BWR core design

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

    International Nuclear Information System (INIS)

    Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

    1984-01-01

    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.) [de

  7. Development of a real-time stability measurement system for boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; King, W.T.

    1987-01-01

    This paper describes the development of a portable, real time system for boiling water reactor (BWR) stability measurements. The system provides a means for the operator to monitor the reactor stability using existing plant instrumentation and commercially available hardware. The noise component (i.e., perturbations around steady state) of the neutron signal in BWRs has been shown to contain information about reactor stability, and several algorithms have been developed to extract that information. For the present work, the authors have used an algorithm that has been implemented on a portable personal computer. This algorithm uses the autocorrelation function of naturally occurring neutron noise (measured without special plant perturbations) and an autoregressive modeling technique to produce the asymptotic DR. For this real-time implementation, neutron noise data is preconditioned (i.e., filtered and amplified) and sampled at a 5-Hz sampling rate using a commercial data-acquisition system. Approximately every 1.5 min, the current (snapshot) autocorrelation function is computed directly from the data, and the average autocorrelation is updated. The current and average DR estimates are evaluated with the same periodicity and are displayed on the screen along with the autocorrelations and average power spectrum of the neutron noise

  8. Nonlinear dynamics and stability of boiling water reactors: qualitative and quantitative analyses

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1985-01-01

    A phenomenological model has been developed to simulate the qualitative behavior of boiling water reactors (BWRs) in the nonlinear regime under deterministic and stochastic excitations. After the linear stability threshold is crossed, limit cycle oscillations appear due to interactions between two unstable equilibrium points and the phase-space trajectories. This limit cycle becomes unstable when the feedback gain exceeds a certain critical value. Subsequent limit cycle instabilities produce a cascade of period-doubling bifurcations that leads to a periodic pulsed behavior. Under stochastic excitations, BWRs exhibit a single characteristic resonance, at approx.0.5 Hz, in the linear regime. By contrast, this work shows that harmonics of this characteristic frequency appear in the nonlinear regime. Furthermore, this work also demonstrates that amplitudes of the limit cycle oscillations do not depend on the variance of the stochastic excitation and remain bounded at all times. A physical model of nonlinear BWR dynamics has also been developed and employed to calculate the amplitude of limit cycle oscillations and their effects on fuel integrity over a wide range of operating conditions in the Vermont Yankee reactor. These calculations have confirmed that, beyond the threshold for linear stability, the reactor's state variable undergo limit cycle oscillations

  9. Fact and fiction in ECP measurement and control in boiling water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Macdonald, D.D.

    2005-01-01

    A review is presented of various electrochemical potentials, including the electrochemical corrosion potential (ECP), that are used in the mitigation of stress corrosion cracking in the primary coolant circuits of boiling water reactors (BWRs). Attention is paid to carefully defining each potential in terms of fundamental electrochemical concepts, so as to counter the confusion that has arisen due to the misuse of previously accepted terminology. A brief discussion is also included of reference electrodes and it is shown on the basis of experimental data that the use of a platinum redox sensor as a reference electrode in the monitoring of ECP in BWR primary coolant circuits is inappropriate and should be discouraged. If platinum is used as a reference electrode, because of extenuating circumstances (e.g., potential measurements in high dose regions in a reactor core), the onus must be placed on the user to demonstrate quantitatively that the electrode behaves as an equilibrium electrode under the specified conditions and/or that its potential is invariant with changes in the independent variables of the system. Preferably, a means should also be demonstrated of transferring the measured potential to the standard hydrogen electrode (SHE) scale. (orig.)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Gerstner, Douglas M.

    2009-01-01

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

  12. To the analysis of reactor noise in boiling water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

    The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

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

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

    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)

  14. Fault tolerant digital control systems for boiling water reactors

    International Nuclear Information System (INIS)

    Chakraborty, S.; Cash, N.R.

    1986-01-01

    In a Boiling Water Reactor nuclear power plant, the power generation control function is divided into several systems, each system controlling only a part of the total plant. Presently, each system is controlled by conventional analog or digital logic circuits with little interaction for coordinated control. The advent of microprocessors has allowed the development of distributed fault-tolerant digital controls. The objective is to replace these conventional controls with fault-tolerant digital controls connected together with digital communication links to form a fully integrated nuclear power plant control system

  15. Domestic and overseas development of advanced boiling water reactors

    International Nuclear Information System (INIS)

    Hatazawa, Mamoru; Fuchino, Satoshi; Nakada, Kotaro

    2010-01-01

    Since Toshiba delivered the world's first advanced boiling water reactor (ABWR) to The Tokyo Electric Power Company, Inc. in 1996, we have been devoting continuous efforts to the construction and operational support of ABWR systems as major products. We are now promoting the construction of domestic and overseas ABWR systems along with the standardization of ABWRs. We are also engaged in the research and development of core technologies to support further promotion of ABWRs as a concurrent solution to the issues of global warming and energy security for individual countries. (author)

  16. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  17. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    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)

  18. Modern technology applied in the advanced BWR (ABWR)

    International Nuclear Information System (INIS)

    Hucik, S.A.

    1988-01-01

    The advanced boiling water reactor (ABWR) represents the next generation of light water reactors (LWR) to be introduced into commercial operation in the 1990's. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. This paper discusses how the ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. A technical evaluation of the ABWR shows its superiority in terms of performance characteristics and economics relative to current LWR designs

  19. Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mkrtchyan, Lilit; Schau, Henry [TUEV SUED Energietechnik GmbH, Mannheim (Germany). Abt. Strukturverhalten; Wolf, Werner; Holzer, Wieland [TUEV SUED Industrie Service GmbH, Muenchen (Germany). Abt. Behaelter und Turbosatz; Wernicke, Robert; Trieglaff, Ralf [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany). Abt. Festigkeit und Konstruktion

    2011-08-15

    The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

  20. Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

    International Nuclear Information System (INIS)

    Mkrtchyan, Lilit; Schau, Henry; Wolf, Werner; Holzer, Wieland; Wernicke, Robert; Trieglaff, Ralf

    2011-01-01

    The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

  1. Advanced analytical techniques for boiling water reactor chemistry control

    Energy Technology Data Exchange (ETDEWEB)

    Alder, H P; Schenker, E [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-02-01

    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.

  2. Numerical evaluation of fluid mixing phenomena in boiling water reactor using advanced interface tracking method

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Takase, Kazuyuki

    2008-01-01

    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. (author)

  3. Seismic evaluation of BWR spent fuel storage racks using actual damping by vibration test in water

    International Nuclear Information System (INIS)

    Yamasaki, Hiroto; Iwakura, Shigeyoshi; Imaoka, Tetsuo; Okumura, Kazue; Orita, Syuichi; Namita, Yoshio

    2010-01-01

    Damping value for BWR spent fuel storage racks has been used 1 percent damping, which is applied to welded steel structures in air as defined JEAG4601. However, it is considered that the actual damping is higher than that of the above mentioned, because of its underwater installation. This report shows the actual damping value of the Check Arrayed Rack by vibration test in water and Evaluation by the analysis of rack using actual damping. (author)

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

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

    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)

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

    International Nuclear Information System (INIS)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  8. 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.

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  10. Method of providing extended life expectancy for components of boiling water reactors

    International Nuclear Information System (INIS)

    Niedrach, L.W.

    1992-01-01

    This patent describes a containment for a boiling water nuclear reactor, a stainless steel containment, the containment having a deposit of a metal of the platinum group of metals on the surfaces thereof exposed to high temperature, high pressure water of the boiling water nuclear reactor

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

    International Nuclear Information System (INIS)

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

    1980-07-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  14. 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)

  15. 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)

  16. BWR shutdown analyzer using artificial intelligence (AI) techniques

    International Nuclear Information System (INIS)

    Cain, D.G.

    1986-01-01

    A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

  17. Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

    International Nuclear Information System (INIS)

    Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

    2012-01-01

    Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

  18. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  19. Noise analysis of the Dodewaard boiling water reactor: characteristics and time history

    International Nuclear Information System (INIS)

    Veer, J.H.C. v.d.; Kema, N.V.

    1982-01-01

    Reactor noise measurements have been performed in the Dodewaard BWR since the eighth fuel cycle (1978). Analysis of the noise characteristics of the ex-core neutron detectors are reported. As a result characteristics of the global component of the boiling noise and the influence of oscillatory effects in reactor pressure control and steam flow rate are described. The influence of power feedback effects on the detection of global noise at low frequencies is given using point kinetic reactor theory. Results are reported on the behaviour of the neutron noise characteristics during one fuel cycle and on the behaviour from fuel cycle 8 to 11. (author)

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

    International Nuclear Information System (INIS)

    Fukutomi, S.; Takigawa, Y.

    1992-01-01

    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

  1. Boiling water reactor liquid radioactive waste processing system

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

  2. Expert system for control rod programming of boiling water reactors

    International Nuclear Information System (INIS)

    Fukuzaki, T.; Yoshida, K.; Kobayashi, Y.; Matsuura, H.; Hoshi, K.

    1986-01-01

    Control rod programming, one of the main tasks in reactor core management of boiling water reactors (BWRs), can be successfully accomplished by well-experienced engineers. By use of core performance evaluation codes, their knowledge plays the main role in searching through optimal control rod patterns and exposure points for adjusting notch positions and exchanging rod patterns. An expert system has been developed, based on a method of knowledge engineering, to lighten the engineer's load in control rod programming. This system utilizes an inference engine suited for planning/designing problems, and stores the knowledge of well-experienced engineers in its knowledge base. In this report, the inference engine, developed considering the characteristics of the control rod programming, is introduced. Then the constitution and function of the expert system are discussed

  3. Boiling water reactor stability analysis in the time domain

    International Nuclear Information System (INIS)

    Borkowski, J.A.

    1991-01-01

    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

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Yadigaroglu, G. [Institute of Energy Technology, Swiss Federal Institute of Technology, ETH, ETH-Zentrum, CLT-C1, CH-8092 Zurich (Switzerland)]. E-mail: yadi@ethz.ch; Askari, B. [Institute of Energy Technology, Swiss Federal Institute of Technology, ETH, ETH-Zentrum, CLT-C1, CH-8092 Zurich (Switzerland)

    2005-05-01

    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.

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

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Askari, B.

    2005-01-01

    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

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

    International Nuclear Information System (INIS)

    Alder, H.P.; Buckley, D.; Grauer, R.; Wiedemann, K.H.

    1989-09-01

    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

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

    International Nuclear Information System (INIS)

    Sakai, Takeshi; Tokunaga, Katsumi; Stevens, G.L.; Ranganath, S.

    1993-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  10. Mitigation of stress corrosion cracking in boiling water reactors

    International Nuclear Information System (INIS)

    Hanneman, R.E.; Cowan, R.L. II

    1980-01-01

    Intergranular stress corrosion cracking (IGSCC) has occurred in a statistically small number of weld heat affected zones (HAZ) of 304 SS piping in BWR's. A range of mitigating actions have been developed and qualified that provide viable engineering solutions to the unique aspects of (1) operating plants, (2) plants under various stages of construction, and (3) future plants. This paper describes the technical development of each mitigating concept, relates it to the fundamental causal factors for IGSCC, and discusses its applicability to operating, in-construction and new BWR's. 31 refs

  11. Calculation model for predicting concentrations of radioactive corrostion products in the primary coolant of boiling water reactors

    International Nuclear Information System (INIS)

    Uchida, S.; Kikuchi, M.; Asakura, Y.; Yusa, H.; Ohsumi, K.

    1978-01-01

    A calculation model was developed to predict the shutdown dose rate around the recirculation pipes and their components in boiling water reactors (BWRs) by simulating the corrosion product transport in primary cooling water. The model is characterized by separating cobalt species in the water into soluble and insoluble materials and then calculating each concentration using the following considerations: (1) Insoluble cobalt (designated as crud cobalt is deposited directly on the fuel surface, while soluble cobalt (designated as ionic cobalt) is adsorbed on iron oxide deposits on the fuel surface. (2) Cobalt-60 activated on the fuel surface is dissolved in the water in an ionic form, and some is released with iron oxide as crud. The model can follow the reduction of 60 Co in the primary cooling water caused by the control of the iron feed rate into the reactor, which decreases the iron oxide deposits on the fuel surface and then reduces the cobalt adsorption rate. The calculated results agree satisfactorily with the measurements in several BWR plants

  12. Auxiliary water supply device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

  13. Seismic risk assessment of a BWR: status report

    International Nuclear Information System (INIS)

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

    1985-02-01

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

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

    International Nuclear Information System (INIS)

    Stosic, Zoran

    1999-01-01

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

  15. Acoustic emissions of a boiling liquid - an experimental survey in water and extrapolation to SFRs

    International Nuclear Information System (INIS)

    Vanderhaegen, M.; Paumel, K.; Tourin, A.

    2013-06-01

    The acoustic detection of sodium boiling is seen as a promising and innovative surveillance technique for sodium-cooled fast reactors (SFRs). It could be especially useful to detect in-core boiling that are the consequence of initiating accidents or whilst the mean subassembly temperature is very close to the nominal value. This latter is a consequence of the fuel assembly design of SFRs. Furthermore, it is a technique that has been proven to be successful in the past to follow the boiling behavior during SFR experiments that were aimed at simulating accidental conditions. However its effectiveness as in-core instrumentation still has to be demonstrated. In that aim, the acoustic emissions of sodium boiling in subassemblies are studied. Experimental studies are however limited to the boiling of common coolants due to the complications that arise when boiling liquid metals. As such, simple water experiments are performed. And although the results of these experiments are not completely representative for sodium boiling due to the incomplete thermo-hydraulic similarities between sodium and water, they can provide an interesting knowledge of the many influences that control the acoustic pressure field. In this article we'll specifically show how the condensation of vapor in subcooled liquid, the principal contribution to the acoustic emissions during boiling and hence the acoustic spectrum, is influenced by a pin-bundle geometry. We study this influence by comparing pool boiling experimental acoustic recordings with those of a simple pin-bundle geometry boiling experiment. The qualitative link, between this relatively simple pin-bundle experiment and the condensation phenomena that take place during sodium boiling inside SFR subassemblies, is used as a basis for this analysis. This simple experimental evidence, together with other theoretical arguments based on a thorough analysis of the sodium material properties, enables us to deduce that simple sodium

  16. The analogy between the bubbling of air into water and nucleate boiling at saturation temperature

    International Nuclear Information System (INIS)

    Wallis, G.B.

    1960-01-01

    This paper presents a case for the separate consideration of the hydrodynamic and thermal aspects of nucleate boiling. It is shown how boiling phenomena may be simulated in detail by the use of porous media to introduce air bubbles into water. Points of similarity and equivalence are described and analysed. (author)

  17. 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.

  18. Identification of neutron noise sources in a boiling water reactor

    International Nuclear Information System (INIS)

    Sides, W.H. Jr.; Mathis, M.V.; Smith, C.M.

    1977-01-01

    Measurements were made at units 2 and 3 of the Browns Ferry Nuclear Power Plant in order to characterize the noise signatures of the neutron and process signals and to determine the usefulness of such signatures for anomaly detection in BWR-4s. Previous measurements and theoretical analyses of BWR noise by others were concerned with the determination of steam velocity and void fraction (using the local component of neutron noise) and with the sources of global noise. The work described is under a five-part program to develop a complete and systematic analysis and representation of BWR neutron and process noise through complementary measurements and stochastic model developments. The parts are: (1) recording as many neutron detector and process noise signals as are available in a BWR-4; (2) reducing these data to noise signatures in order to perform an empirical analysis of these signatures, and documenting the relationships between the signals from spatially separated neutron detectors and between neutron and process variables; (3) developing spatially dependent neutronic models coupled with thermal-hydraulic models to aid in interpreting the observed relationships among the measured noise signatures, (4) comparing measured noise signatures with model predictions to obtain additional insight into BWR-4 dynamic behavior and to validate the models; and (5) using these models to predict the sensitivity of noise monitoring for detection, surveillance, and diagnosis of postulated in-core anomalies in BWRs. The paper describes the procedures used to obtain the noise recordings and presents initial empirical analysis and observations pertaining to the noise signatures and the relationships between several noise variables in the 0.01- to 1-Hz range. The mathematical models have not been developed sufficiently to report theoretical results or to compare measured spectra with model predictions at this time

  19. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  20. Critical discharge of initially subcooled water through slits. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N; Schrock, V E

    1983-09-01

    This report describes an experimental investigation into the critical flow of initially subcooled water through rectangular slits. The study of such flows is relevant to the prediction of leak flow rates from cracks in piping, or pressure vessels, which contain sufficient enthalpy that vaporization will occur if they are allowed to expand to the ambient pressure. Two new analytical models, which allow for the generation of a metastable liquid phase, are developed. Experimental results are compared with the predictions of both these new models and with a Fanno Homogeneous Equilibrium Model.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  2. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  3. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Miki, Minoru; Ohki, Arahiko.

    1984-01-01

    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)

  5. Study and application of boiling water reactor jet pump characteristic

    International Nuclear Information System (INIS)

    Liao Lihyih

    1992-01-01

    RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model. In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible. In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power, level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K d , the suction flow junction form loss coefficients, K s , the diffuser form loss coefficient, K c , and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve. (orig.)

  6. Recent computer applications in boiling water reactor power plants

    International Nuclear Information System (INIS)

    Hiraga, Shoji; Joge, Toshio; Kiyokawa, Kazuhiro; Kato, Kanji; Nigawara, Seiitsu

    1976-01-01

    Process computers in boiling water reactor power plants have won the position of important equipments for the calculation of the core and plant performances and for data logging. Their application technique is growing larger and larger every year. Here, two systems are introduced; plant diagnostic system and computerized control panel. The plant diagnostic system consists of the part processing the signals from a plant, the operation part mainly composed of a computer to diagnose the operating conditions of each system component using input signal, and the result display (CRT or typewriter). The concept on the indications on control panels in nuclear power plants is changing from ''Plant parameters and to be indicated on panel meters as much as possible'' to ''Only the data required for operation are to be indicated.'' Thus the computerized control panel is attracting attention, in which the process computer for processing the operating information and CRT display are introduced. The experimental model of that panel comprises and operator's console and a chief watchmen's console. Its functions are dialogic data access and the automatic selection of preferential information. (Wakatsuki, Y.)

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

    International Nuclear Information System (INIS)

    Prasad, Y.S.R.; Srinivasan, V.S.

    1994-01-01

    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

  8. Generic risk insights for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    Travis, R.; Taylor, J.; Chung, J.

    1991-05-01

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

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

    International Nuclear Information System (INIS)

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

    1995-03-01

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

  10. The development of fuel elements for boiling water reactors

    International Nuclear Information System (INIS)

    Holzer, R.; Kilian, P.

    1984-01-01

    The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Dan, Ho Jin; Lee, Joon Sik

    2016-01-01

    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

  13. 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.

  14. Effects of torus wall flexibility on forces in the Mark I Boiling Water Reactor Pressure Suppression System. Part I

    International Nuclear Information System (INIS)

    Martin, R.W.; McCauley, E.W.

    1977-09-01

    The authors investigated the effects of torus wall flexibility in the pressure suppression system of a Mark I boiling water reactor (BWR) when the torus wall is subjected to hydrodynamic loadings. Using hypothetical models, they examined these flexibility effects under two hydrodynamic loading conditions: (1) a steam relief valve (SRV) discharge pulse, and (2) a loss-of-coolant accident (LOCA) chugging pulse. In the analyses of these events they used a recently developed two-dimensional finite element computer code. Taking the basic geometry and dimensions of the Monticello Mark I BWR nuclear power plant (in Monticello, Minnesota, U.S.A.), they assessed the effects of flexibility in the torus wall by changing values of the inside-diameter-to-wall-thickness ratio. Varying the torus wall thickness (t) with respect to the inside diameter (D) of the torus, they assigned values to the ratio D/t ranging from 0 (infinitely rigid) to 600 (highly flexible). In the case of a modeled steam relief valve (SRV) discharge pulse, they found the peak vertical reaction force on the torus was reduced from that of a rigid wall response by a factor of 3 for the most highly flexible, plant-simulated wall (D/t = 600). The reduction factor for a modeled loss-of-coolant accident (LOCA) chugging pulse was shown to be 1.5. The two-dimensional analyses employed overestimate these reduction factors but have provided, as intended, definition of the effect of torus boundary stiffness. In the work planned for FY79, improved modeling of the structure and of the source is expected to result in factors more directly applicable to actual pressure suppression systems

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

    International Nuclear Information System (INIS)

    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

  16. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. State of development progress of advanced boiling water reactors

    International Nuclear Information System (INIS)

    Tomono, Katsuya

    1982-01-01

    Advanced BWRs being developed at present are those aiming at the improvement of reliability and safety, the reduction of radiation exposure, the improvement of operation performance and capacity ratio of plants, and the heightening of economical efficiency by concentrating the experience and excellent technology of BWR manufacturers in the world. Now in Japan, the independence with Japanese technology is possible in almost all fields of nuclear power generation, and the improvement and standardization project is in progress to obtain the steady results. However, in order to pursue the most desirable BWRs conceivable at present, five BWR manufacturers in the world organized the Advanced Engineering Team in July, 1978, and performed the feasibility study of advanced BWRs for more than one year. Tokyo Electric Power Co., Inc., evaluated the report on the results, and judged that it is desirable to advance into the next stage aiming at the practical use of advanced BWRs. For the purpose, the electric power common research on advanced BWRs has been in progress, and the A-BWR project is to be examined in the third improvement and standardization project of MITI. The main technical features such as the coolant recirculation system of internal pump type, reinforced concrete containment vessels, fine motion control rod drive, improved core and fuel and others are explained. (Kako, I.)

  18. Complete BWR--EM LOCA analysis using the WRAP--EM system

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

    1979-01-01

    The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

  19. An analysis of the water-level monitoring system for a boiling-water reactor

    International Nuclear Information System (INIS)

    Carlson, R.W.; Belblidia, L.A.; Russell, J.L. Jr.

    1985-01-01

    The water-level instrumentation system is very important to the overall safety of a BWR. This system is being monitored by the Safety Parameter Display System (SPDS) that is being installed in Georgia Power Company's Plant Hatch. One of the most significant functions of the SPDS is the comparison of redundant instrument readings and formation of the best estimate of each parameter from those readings which are consistent. When comparing water-level instrument readings, it is necessary to correct the individual readings for differences between current and calibration conditions as well as for differences between calibration conditions for the multiple instruments. This paper documents the examination of the water-level instrumentation system at Plant Hatch and presents the development of the equations that were used to determine the differences between indicated and actual water levels. (author)

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

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shung-Jung; Kuo, Weng-Sheng

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Edwards, Robert; Huang, Zhengyu

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Ott, L.J.

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Tsuiki, Makoto; Sakurada, Koichi; Yoshida, Hiroyuki.

    1988-01-01

    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)

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

    International Nuclear Information System (INIS)

    Xiaodi, Xue; Hongfei, Zheng; Kaiyan, He; Zhili, Chen; Tao, Tao; Guo, Xie

    2010-01-01

    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/m 2 , the boiled water yield rate of the system has reached 20 kg/h and its total energy efficiency has exceeded 40%.

  5. Incorporating Water Boiling in the Numerical Modelling of Thermal Remediation by Electrical Resistance Heating

    Science.gov (United States)

    Molnar, I. L.; Krol, M.; Mumford, K. G.

    2017-12-01

    Developing numerical models for subsurface thermal remediation techniques - such as Electrical Resistive Heating (ERH) - that include multiphase processes such as in-situ water boiling, gas production and recovery has remained a significant challenge. These subsurface gas generation and recovery processes are driven by physical phenomena such as discrete and unstable gas (bubble) flow as well as water-gas phase mass transfer rates during bubble flow. Traditional approaches to multiphase flow modeling soil remain unable to accurately describe these phenomena. However, it has been demonstrated that Macroscopic Invasion Percolation (MIP) can successfully simulate discrete and unstable gas transport1. This has lead to the development of a coupled Electro Thermal-MIP Model2 (ET-MIP) capable of simulating multiple key processes in the thermal remediation and gas recovery process including: electrical heating of soil and groundwater, water flow, geological heterogeneity, heating-induced buoyant flow, water boiling, gas bubble generation and mobilization, contaminant mass transport and removal, and additional mechanisms such as bubble collapse in cooler regions. This study presents the first rigorous validation of a coupled ET-MIP model against two-dimensional water boiling and water/NAPL co-boiling experiments3. Once validated, the model was used to explore the impact of water and co-boiling events and subsequent gas generation and mobilization on ERH's ability to 1) generate, expand and mobilize gas at boiling and NAPL co-boiling temperatures, 2) efficiently strip contaminants from soil during both boiling and co-boiling. In addition, a quantification of the energy losses arising from steam generation during subsurface water boiling was examined with respect to its impact on the efficacy of thermal remediation. While this study specifically targets ERH, the study's focus on examining the fundamental mechanisms driving thermal remediation (e.g., water boiling) renders

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

    International Nuclear Information System (INIS)

    Tomita, M.; Haruyama, Y.; Saito, M.

    1993-01-01

    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 cm 2 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.)

  7. The predicted effectiveness of noble metal treatment at the Chinshan boiling water reactor

    International Nuclear Information System (INIS)

    Yeh Tsungkuang; Chu Fang; Chang Ching; Huang Chiashen

    2000-01-01

    The technique of noble metal treatment (NMT) available in a form of noble metal cooling (NMC) or noble metal chemical addition (NMCA), was introduced to enhance effectiveness of hydrogen water chemistry. Since it is technically difficult to gain access to an entire primary heat transport circuit (PHTC) of a BWR and monitor variation on electrochemical corrosion potential (ECP), a question whether the NMC technology is indeed effective for lowering the ECP of every location in a BWR is not still well understood at the moment. Then, computer modeling is so far the best tool to help investigate effectiveness of the NMT along PHCT of the BWR. Here was discussed on how the computer model was calibrated by using measured chemistry data obtained from No. 2 unit (BWR) in the Kuosheng Plant. The effect of noble metal treatment coupled with hydrogen water chemistry has been quantitatively molded, on a base of two different sets of ECD enhancement data. It was predicted that No. 1 unit in the Chinshan could be protected by noble metal treatment with lower [H 2 ] FW . In the case of competitive enhancing factors for the ECDs of oxygen reduction, hydrogen peroxide reduction, and hydrogen oxidation reactions, HWC had always to be present for noble metal treatment to be effective for protecting a reactor. Otherwise, according to a model calculation based upon the results from Kim's work, the ECP might instead be increased due to the enhanced reduction reaction rate of oxygen and hydrogen peroxide, especially in the near core regions. (G.K.)

  8. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

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

  9. Activity buildup in the recirculation system of boiling water reactors

    International Nuclear Information System (INIS)

    Grauer, R.

    1987-05-01

    The deposition of activated corrosion products in the recirculation system of boiling water reactors produces increased radiation levels which lead to a corresponding increase in personnel irradiation dose rates during shut down and maintenance. The major part of this dose rate is due to Co-60. Based on a comprehensive literature study concerning this theme, the author has attempted to identify the individual stages of the activity build up and to classify their importance. The following areas are discussed in detail: - the origins of the corrosion products and of Co-59 in the reactor feedwater; - the consolidation of the cobalt in the fuel element crud deposits (activation); - the release and transport of the Co-60; - the build up of Co-60 in the corrosion products of the primary circuit. Existing models for the build up of cirquit radioactivity are discussed and the operating experience from selected reactors are summarized. Finally the state of the art of knowledge concerning the individual stages in the development of the activation build up is depicted. This highlights the existing gaps and thus identifies areas for possible R+D activities. Corrosion chemistry aspects of the cobald build up in the primary cirquit have already been studied on a broad basis and are continuing to be researched in a number of centers. The crystal chemistry of austenitic steel corrosion products poses a number of yet unanswered questions. There are major loopholes associated with an understanding of the activation processes of the cobalt deposited on the fuel elements and in the mass transfer of Co-60. For these processes, the most important influence stems from factors associated with colloid chemistry. In this respect there is ample room for new and original research contributions. It is recommended that sections 8 and 9 of this report should be read as an 'Executive Summary'. (author)

  10. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

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

  11. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

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

  12. TARMS, an on-line boiling water reactor operation management system. [3 D core simulator LOGOS 2

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

    1984-12-01

    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.

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

    International Nuclear Information System (INIS)

    Lin Chaung; Lin, Tung-Hsien

    2012-01-01

    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.

  14. Comparison of finite element and influence function methods for three-dimensional elastic analysis of boiling water reactor feedwater nozzle cracks. Phase report

    International Nuclear Information System (INIS)

    Besuner, P.M.; Caughey, W.R.

    1976-11-01

    The finite element (FE) and influence function (IF) methods are compared for a three-dimensional elastic analysis of postulated circular-shaped surface cracks in the feedwater nozzle of a typical boiling water reactor (BWR). These are two of the possible methods for determining stress intensity factors for nozzle corner cracks. The FE method is incorporated in a direct manner. The IF method is used to compute stress intensity factors only when the uncracked stress field (i.e., the stress in the uncracked solid at the locus of the crack to be eventually considered) has been computed previously. Both the IF and FE methods are described in detail and are applied to several test cases chosen for their similarity to the nozzle crack problem and for the availablility of an accurate published result obtained from some recognized third method of solution

  15. 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)

    Demaziere, C.

    1999-02-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. 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

  16. Water chemistry experiences in a Japanese BWR during preoperational test period

    International Nuclear Information System (INIS)

    Nagasawa, K.; Maeda, K.; Nakamura, M.

    1994-01-01

    The interest in organic impurities intrusion into the BWR's coolant has been increased in recent years. Organic impurities can be measured as total organic carbons (TOC) by using commercial TOC measuring instruments. The corrosive impurities such as sulfates and chlorides, produced by decomposition of the TOC, will enhance the crack growth rate of stress corrosion cracking (SCC) in the reactor cooling system. During the preoperational test period, the reactor water quality changed remarkably by TOC intrusion. The TOC intrusion was monitored by measuring main steam conductivity. It is appeared that the main steam conductivity measurement is one of the effective and reliable techniques to detect the TOC intrusion into the reactor. (authors). 4 figs., 5 refs

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  18. 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)

  19. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  20. Single-phase flow and flow boiling of water in horizontal rectangular microchannels

    OpenAIRE

    Mirmanto

    2013-01-01

    This thesis was submitted for the degree of Doctor of Philosophy and awarded by Brunel University The current study is part of a long term experimental project devoted to investigating single-phase flow pressure drop and heat transfer, flow boiling pressure drop and heat transfer, flow boiling instability and flow visualization of de-ionized water flow in microchannels. The experimental facility was first designed and constructed by S. Gedupudi (2009) and in the present study; ...

  1. Critical heat flux for flow boiling of water in mini-channels

    International Nuclear Information System (INIS)

    Zhang, Weizhong; Mishima, Kaichiro; Hibiki, Takashi

    2007-01-01

    Critical heat flux (CHF) is a limiting factor when flow boiling is applied to dissipate high heat flux in mini-channels. In view of practical importance of critical heat flux correlations in engineering design and prediction, this study presents an evaluation of existing CHF correlations for flow boiling of water with available databases taken from small-diameter tubes, and then develops a new, simple CHF correlation for flow boiling in mini-channel. Three correlations by Bowring, Katto and Shah are evaluated with available CHF data in the literature for saturated flow boiling, and three correlations by Inasaka-Nariai, Celata et al. and Hall-Mudawar evaluated with the CHF data for subcooled flow boiling. The Hall-Mudawar correlation and the Shah correlation appear to be the most reliable tools for CHF prediction in the subcooled and saturated flow boiling regions, respectively. In order to avoid the defect of predictive discontinuities often encountered when applying previous correlations, a simple, nondimensional, inlet conditions dependent CHF correlation for saturated flow boiling has been formulated. Its functional form is determined by application of the artificial neural network and parametric trend analyses to the collected database. Superiority of this new correlation has been verified by the collected database. It has a mean deviation of 16.8% for this collected databank, smallest among all tested correlations. Compared to many inordinately complex correlations, this new correlation consists only of one single equation. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Effect of water impurities on stress corrosion cracking in a boiling water reactor

    International Nuclear Information System (INIS)

    Ljungbery, L.G.; Cubicciotti, D

    1985-01-01

    A series of stress corrosion tests, including corrosion potential and water chemistry measurements, has been performed in the Swedish Ringhals-1 boiling water reactor. Tests have been run under reactor start-up and reactor power operation with normal reactor water conditions and with alternate water chemistry in which hydrogen is added to the feedwater to suppress stress corrosion cracking. During one alternate water chemistry test, there was significant intergranular corrosion cracking of sensitized stainless specimens. It is shown that nitrate and sulfate, arising from an accidental resin intrusion, are likely causes. Nitrate increases the oxidizing power of the water, and sulfate enhances cracking under oxidizing conditions. During another test under start-up conditions, enhanced transgranular stress corrosion cracking in low alloy steels and possibly initiation of cracking in a nickel base alloy was observed as a result of resin intrusion into the reactor water. The intrusion produced acid and sulfate, which are believed to enhance hydrogen cracking conditions

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

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  5. Water chemistry experience following an extensive power up-rate in Oskarshamn 3 BWR

    International Nuclear Information System (INIS)

    Wegemar, Boerje; Nilsson, Jimmy; Lejon Johan; Bergfors, Asa; Arnberg, Bo

    2012-09-01

    The Swedish Oskarshamn 3 BWR plant, operated by OKG, was first connected to the grid in 1985. The plant has been power up-rated in two steps; from the original design, 3020 MWth, to 3300 MWth (109%, 1989) and recently to 3900 MWth (129%, 2009). Westinghouse Electric Sweden AB (former ASEA-Atom, OEM of the plant) was rewarded a major contract in the recently implemented up-rating project, the PULS project. The PULS project is quite unique since no operating experience has to date been reported from a similar major power up-rate in a BWR plant. Water chemistry experience from the first period of operation following the implementation of the PULS project is reported and discussed in the paper. Reported chemistry and radiochemistry measurements in feedwater (FW) and reactor water (RW) include corrosion products, activated corrosion products, dissolved oxygen and impurities like chloride, sulfate etc. Furthermore, a comparison of water quality prior to implementation of the PULS project is included. Several process systems have been modified, one of them being the condensate cleanup system (CCU), a Pre-coat filter system. The design criteria for the CCU system include the filter run-lengths, pressure drop before back-washing and requirements on water chemistry quality. The paper describes in some detail the CCU system modifications being implemented in order to fulfil the design criterion. CCU cleanup efficiency, operating temperature and influence of hydrogen peroxide on the CCU resin are all important issues being covered in the paper. As for the latter, it is well known that oxygen and hydrogen peroxide (from radiolysis in the core region) might cause partial deterioration of CCU standard cation resin resulting in increased RW sulfate concentrations. This aspect is covered in the paper as well. The reactor water cleanup system (RWCU) in Oskarshamn 3 consists of deep bed ion exchange filters (mixed bed filter). The purpose of RWCU is to maintain a low level of

  6. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

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

  7. 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)

  8. Decontamination flange film characterization for a boiling water reactor under hydrogen water chemistry

    International Nuclear Information System (INIS)

    Baston, V.F.; Garbauskas, M.F.; Bozeman, J.

    1996-01-01

    Stainless steel artifacts removed from a boiling water reactor class 4 plant that operated under hydrogen water chemistry and experienced a difficult decontamination were submitted for oxide film characterization. The results reported for the corrosion film composition and structure are consistent with existing theoretical concepts for stainless steel corrosion, spinel structure site preferences (octahedral or tetrahedral) for transition metal ions, and potential-pH diagrams. The observed zinc effects on film stability and lower cobalt incorporation are also consistent with these theoretical concepts

  9. Generic safety insights for inspection of boiling water reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

    As the number of operating nuclear power plants (NPP) increases, safety inspection has increased in importance. However, precisely what is important, and what is not important? What should one focus inspection efforts on. Over the last two years Probabilistic Risk Assessment (PR) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only about 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the BNL program to develop generic BWR PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs

  10. Pressure suppression containment system for boiling water reactor

    Science.gov (United States)

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

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

  11. Early hydrogen water chemistry in the boiling water reactor: industry-first demonstration

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Odell, Andrew D.; Giannelli, Joseph F.

    2012-09-01

    Hydrogen injection into the BWR feedwater during power operation has resulted in significant IGSCC reductions. Further, noble metal application (NMCA) during shutdown or On-line NobleChem TM (OLNC) during power operation has greatly reduced the required hydrogen injection rate by catalyzing the hydrogen-oxygen reaction on the metal surfaces, reducing the electrochemical corrosion potential (ECP) at operating temperature to well below the mitigation ECP of -230 mV (SHE) at reactor water hydrogen to oxidant (O 2 + H 2 O 2 ) molar ratios of ≥2. Since IGSCC rates increase markedly at reduced temperature, and the potential for crack initiation exists, additional crack mitigation was desired. To close this gap in mitigation, the EPRI BWR Startup ECP Reduction research and development program commenced in 2008 to undertake laboratory and feasibility studies for adding a reductant to the reactor water system during start-ups. Under this program, ECP reductions of noble metal treated stainless steel sufficient to mitigate IGSCC at startup temperatures were achieved in the laboratory in the absence of radiation at hydrogen, hydrazine and carbohydrazide to oxygen molar ratios of ≥ 2, ≥1.5 and ≥0.7, respectively. Based on the familiarity of operating BWRs with using hydrogen, a demonstration of hydrogen injection during the startup of an actual BWR using noble metals was planned. This process, named EHWC (Early Hydrogen Water Chemistry), differs from the HDS (Hydrogen During Startup) approach that has been successful in Japan in that HDS injects sufficient hydrogen for bulk oxidant reduction whereas EHWC injects a smaller amount of hydrogen, sufficient to achieve a hydrogen:oxidant molar ratio of at least two at noble metal treated surfaces. The industry-first EHWC demonstration was performed at Exelon's Peach Bottom 3 nuclear power plant in October 2011. Prior to EHWC, Peach Bottom 3 had one NMCA (October 1999) and five annual OLNC applications (starting in 2007

  12. Experimental investigation of mixing of non-isothermal water streams at BWR operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bergagio, Mattia, E-mail: bergagio@kth.se [AlbaNova University Center, Nuclear Reactor Technology Division, Department of Physics, Royal Institute of Technology, 106 91 Stockholm (Sweden); Anglart, Henryk, E-mail: henryk@kth.se [AlbaNova University Center, Nuclear Reactor Technology Division, Department of Physics, Royal Institute of Technology, 106 91 Stockholm (Sweden); Institute of Heat Engineering, Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw (Poland)

    2017-06-15

    Highlights: • Temperatures are measured in the presence of mixing at BWR operating conditions. • The thermocouple support is moved along a pattern to extend the measurement region. • Uncertainty of 1.58 K for temperatures acquired at 1000 Hz. • Momenta of the hot streams and thermal stratification affect the data examined. • Unconventional spectral analysis is required to further study the data collected. - Abstract: In this experimental investigation, wall surface temperatures have been measured during mixing of three water streams in the annular gap between two coaxial stainless-steel tubes. The inner tube, with an outer diameter of 35 mm and a thickness of 5 mm, holds six K-type, ungrounded thermocouples with a diameter of 0.5 mm, which measured surface temperatures with a sampling rate of either 100 Hz or 1000 Hz. The tube was rotated from 0 to 360° and moved in a range of 387 mm in the axial direction to allow measurements of surface temperatures in the whole mixing region. The outer tube has an inner diameter of 80 mm and a thickness of 10 mm to withstand a water pressure of 9 MPa. A water stream at a temperature of either 333 K or 423 K and a Reynolds number between 1657 and 8410 rose vertically in the annular gap and mixed with two water streams at a temperature of 549 K and a Reynolds number between 3.56 × 10{sup 5} and 7.11 × 10{sup 5}. These two water streams entered the annulus radially on the same axial level, 180° apart. Water pressure was kept at 7.2 MPa. Temperature recordings were performed at five axial and eight azimuthal locations, for each set of boundary conditions. Each recording lasted 120 s to provide reliable data on the variance, intermittency and frequency of the surface temperature time series at hand. Thorough calculations indicate that the uncertainty in the measured temperature is of 1.58 K. The mixing region extends up to 0.2 m downward of the hot inlets. In most cases, measurements indicate non-uniform mixing in the

  13. 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)

    Demaziere, C.

    1999-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). 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

  14. 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)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    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

  15. A new coupling kernel for the three-dimensional simulation of a boiling water reactor core by the nodal coupling method

    International Nuclear Information System (INIS)

    Gupta, N.K.

    1981-01-01

    A new coupling kernel is developed for the three-dimensional (3-D) simulation of Boiling Water Reactors (BWR's) by the nodal coupling method. The new kernel depends not only on the properties of the node under consideration but also on the properties of its neighbouring nodes. This makes the kernel more useful in particular for fuel bundles lying in a surrounding of different nuclear characteristics, e.g. for a controlled bundle in the surrounding of uncontrolled bundles or vice-versa. The main parameter in the new kernel is a space-dependent factor obtained from the ratio of thermal-to-fast flux. The average value of the above ratio for each node is evaluated analytically. The kernel is incorporated in a 3-D BWR core simulation program MOGS. As an experimental verification of the model, the cycle-6 operations of the two units of the Tarapur Atomic Power Station (TAPS) are simulated and the result of the simulation are compared with Travelling Incore Probe (TIP) data. (orig.)

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

    International Nuclear Information System (INIS)

    Powers, J.; Yonezawa, H.; Aoyagi, Y.; Kataoka, K.

    2015-09-01

    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. Coarse mesh finite element method for boiling water reactor physics analysis

    International Nuclear Information System (INIS)

    Ellison, P.G.

    1983-01-01

    A coarse mesh method is formulated for the solution of Boiling Water Reactor physics problems using two group diffusion theory. No fuel assembly cross-section homogenization is required; water gaps, control blades and fuel pins of varying enrichments are treated explicitly. The method combines constrained finite element discretization with infinite lattice super cell trial functions to obtain coarse mesh solutions for which the only approximations are along the boundaries between fuel assemblies. The method is applied to bench mark Boiling Water Reactor problems to obtain both the eigenvalue and detailed flux distributions. The solutions to these problems indicate the method is useful in predicting detailed power distributions and eigenvalues for Boiling Water Reactor physics problems

  19. 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)

    Schulz, W.

    2000-01-01

    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)

  20. Nuclear boiling heat transfer and critical heat flux in titanium dioxide-water nanofluids

    International Nuclear Information System (INIS)

    Okawa, Tomio; Takamura, Masahiro; Kamiya, Takahito

    2011-01-01

    Nucleate boiling heat transfer was experimentally studied for saturated pool boiling of water-based nanofluids. Since significant nanoparticle deposition on the heated surface was observed after the nucleate boiling in nanofluids, measurement of CHF was also carried out using the nanoparticle deposited heated surface; pure water was used in the CHF measurement. In the present work, the heated surface was a 20 mm diameter cupper surface, and titanium-dioxide was selected as the material of nanoparticles. Experiments were performed for upward- and downward-facing surfaces. Although the CHFs for the downward-facing surface were generally lower than those for the upward-facing surface, the CHFs for the nanoparticle deposited surface were about 1.9 times greater than those for the bare surface in both the configurations. The CHF improvement corresponded well to the reduction of the surface contact angle. During the nucleate boiling in nanofluids, the boiling heat transfer showed peculiar behavior; it was first deteriorated, then improved, and finally approached to an equilibrium state. This observation indicated that the present nanofluid had competing effects to deteriorate and improve the nucleate boiling heat transfer. It was assumed that the wettability and the roughness of the heated surface were influenced by the deposited nanoparticles to cause complex variation of the number of active nucleation sites. During the nucleate boiling of pure water using the downward-facing surface, a sudden increase in the wall temperature was observed stochastically probably due to the accumulation of bubbles beneath the heated surface. Such behavior was not observed when the pure water was replaced by the nanofluid. (author)

  1. 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

    2016-09-01

    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

  2. The effect of chromate on IGSCC in boiling water reactors - a SSRT study

    International Nuclear Information System (INIS)

    Ullberg, M.

    1992-01-01

    The effect of chromate on IGSCC in Type 304 stainless steel was investigated using the Slow Strain Rate Technique (SSRT). It was found that low concentrations of chromate raises the corrosion potential of SS and causes IGSCC. The effect of chromate was compared to that of the main oxidant in BWRs, hydrogen peroxide. Chromate was found to have less tendency than hydrogen peroxide, at one and the same corrosion potential, to cause IGSCC in SSRT tests. This is interpreted as due to chromate being a better anodic inhibitor than hydrogen peroxide. As a consequence, initiation of IGSCC is slower in the presence of chromate. At least during normal water chemistry, chromate is a secondary oxidant in all of the BWR reactor coolant system. The ECP is then determined by the primary oxidant, hydrogen peroxide. Therefore, the chromate transients which may occur in BWR reactor coolant systems should have no significant effect on IGSCC

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

    International Nuclear Information System (INIS)

    1981-08-01

    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

  4. Sampling system for a boiling reactor NPP

    International Nuclear Information System (INIS)

    Zabelin, A.I.; Yakovleva, E.D.; Solov'ev, Yu.A.

    1976-01-01

    Investigations and pilot running of the nuclear power plant with a VK-50 boiling reactor reveal the necessity of normalizing the design system of water sampling and of mandatory replacement of the needle-type throttle device by a helical one. A method for designing a helical throttle device has been worked out. The quantitative characteristics of depositions of corrosion products along the line of reactor water sampling are presented. Recommendations are given on the organizaton of the sampling system of a nuclear power plant with BWR type reactors

  5. Diffusion bonded matrix of HGMF applied for BWR condensate water purification

    International Nuclear Information System (INIS)

    Soda, Fumitaka; Yukawa, Takao; Ito, Kazuyuki.

    1984-01-01

    High Gradient Magnetic Filter (HGMF) applied to the purification of power plant primary water has recently attracted much attention. In the application of HGMF to the water treatment of power plants, especially nuclear power plants, reliabillties of matrix (filtering medium) as well as removal performance for cruds (insoluble corrosion products) are considered to be important factors. To satisfy these factors, a new filtering medium named Diffision Bonded Matrix (DBM) has been developed and the test results are reported. Filtering efficiency and mechanical stiffness of DBM were examined using HGMF pilot test units consisting of 160 mm diameters x 240 mm length filter. The filtering velocity and the magnetic flux density used in this test were 800 m/h 5 kG, respectively. The filtering efficiencies and of 85-100% were obtained for artificial cruds for DBM. The DBM indicated slightly better filtering efficiency than for conventional wool matrix under the same filtering and matrix conditions. The DBM kept its original mechanical properties and very few pieces of fibers were broken off while the conventional wool matrix lost its volume elasticities and the considerable amount of fibers was broken off during the test operation. The results described here demonstrated the applicability of DBM for treatment of BWR primary water by High Gradient Magnetic Filter. (author)

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

    International Nuclear Information System (INIS)

    Merino C, F.J.; Fuentes C, P.

    2004-01-01

    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)

  7. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

    2006-01-01

    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

  9. 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

  10. 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

  11. Water chemistry guidelines for BWRs

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Jones, R.L.; Welty, C.S.

    1984-01-01

    Guidelines for BWR water chemistry control have been prepared by a committee of experienced utility industry personnel sponsored by the BWR Owners Group on IGSCC Research and coordinated by the Electric Power Research Institute. The guidelines are based on extensive plant operational experience and laboratory research data. The purpose of the guidelines is to provide guidance to the electric utility industry on water chemistry control to help reduce corrosion, especially stress corrosion cracking, in boiling water reactors

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

    Energy Technology Data Exchange (ETDEWEB)

    Sibamoto, Yasuteru [Japan Atomic Energy Research Inst., Tokai Research Establishment, Tokai, Ibaraki (Japan)

    2005-08-01

    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)

  13. Development of alarm handling methods for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yukiharu, Ohga; Hiroshi, Seki; Setsuo, Arita [Power and Industrial Systems R and D Div., Hitachi, Ltd., Hitachi, Ibaraki (Japan)

    1997-09-01

    A method was developed to select important alarms in two steps: first, selection is based on the physical relationship between the alarms, and second, selection is according to the initial event. An approach combining a neural network and knowledge processing was proposed to identify the event rapidly. A prototype system was evaluated in the Kashiwazaki/Kariwa-4 Nuclear Power Plant during the startup test. The evaluation test confirmed that about 30% of the alarms are selected from among the many activated alarms. The second method, dealing with presentation, supports operators in their selection and confirmation of the required information for plant operation. The method selects and offers plant information in response to plant status changes and operators` demands. The selection procedure is based on the knowledge and data as structured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. A prototype system was evaluated using a BWR simulator. The results showed that appropriate information items are automatically selected according to plant status changes and information on generated alarms is presented to operators together with the related trend graph and system diagram. Answers are generated in reply to the operators` demands and operators can confirm the generated alarms on each plant function, such as systems and components. 8 refs, 10 figs, 2 tabs.

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

    International Nuclear Information System (INIS)

    Jimenez Romo, J.A.; Olea Nader, M.A.

    1987-01-01

    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)

  15. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Liu, Maolong; Erkan, Nejdet; Ishiwatari, Yuki; Okamoto, Koji

    2015-01-01

    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

  16. Passive depressurization accident management strategy for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong, E-mail: liuml@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Erkan, Nejdet [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Hitachi-GE Nuclear Energy, Ltd. (Japan); Okamoto, Koji [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan)

    2015-04-01

    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.

  17. Zircaloy spacer grid for boiling light water reactors

    International Nuclear Information System (INIS)

    Borgiani, F.; Cali', G.P.; Cerretti, P.; Pazzo, P.

    1975-01-01

    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

  18. Development of alarm handling methods for boiling water reactors

    International Nuclear Information System (INIS)

    Ohga Yukiharu; Seki Hiroshi; Arita Setsuo

    1997-01-01

    A method was developed to select important alarms in two steps: first, selection is based on the physical relationship between the alarms, and second, selection is according to the initial event. An approach combining a neural network and knowledge processing was proposed to identify the event rapidly. A prototype system was evaluated in the Kashiwazaki/Kariwa-4 Nuclear Power Plant during the startup test. The evaluation test confirmed that about 30% of the alarms are selected from among the many activated alarms. The second method, dealing with presentation, supports operators in their selection and confirmation of the required information for plant operation. The method selects and offers plant information in response to plant status changes and operators' demands. The selection procedure is based on the knowledge and data as structured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. A prototype system was evaluated using a BWR simulator. The results showed that appropriate information items are automatically selected according to plant status changes and information on generated alarms is presented to operators together with the related trend graph and system diagram. Answers are generated in reply to the operators' demands and operators can confirm the generated alarms on each plant function, such as systems and components. 8 refs, 10 figs, 2 tabs

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

    International Nuclear Information System (INIS)

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

    2003-01-01

    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

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

    International Nuclear Information System (INIS)

    Koizumi, Y.; Ohtake, H.; Yuasa, T.; Matsushita, N.

    2001-01-01

    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 X tt ; h TP /h L0 = 5.0(1/ X tt ) 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)

  1. Boiling of water in flow restricted areas modeled by colloidal silica deposits

    International Nuclear Information System (INIS)

    Peixinho, Jorge; Lefevre, Gregory; Coudert, Francois-Xavier; Hurisse, Olivier

    2012-09-01

    Understanding the effects of particle deposits on evaporation and boiling of water represents an important issue for EDF because it causes a severe reduction in efficiency particularly in steam generators of pressurized water reactor. These deposits are made of oxide metallic particles and the deposition process depends on multiple factors. Here we mimic deposits using a simple system made of hydrophilic silica particles. The present study reports experiments on evaporation or boiling of water confined in the pores of colloidal mono-dispersed silica micro-sphere deposits. The boiling of water confined in the pores of the colloidal crystal is studied using optical microscopy, scanning electron microscopy, nitrogen adsorption, water adsorption through infrared attenuated total reflectance spectroscopy, differential scanning calorimetry and thermal gravimetric analysis. By comparison of the results with silica deposits and alumina membranes with cylindrical pores, our study shows that the morphology of the pores contributes to the evaporation and boiling of water. The measurements suggest that particle resuspension and crust formation take place during drying at elevated temperature and are responsible for cracks formation within the deposit film. (authors)

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

    International Nuclear Information System (INIS)

    Kleiss, J.

    1983-01-01

    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.)

  3. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  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. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  6. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  7. Corrosion issues in the BWR and their mitigation for plant life extension

    International Nuclear Information System (INIS)

    Gordon, B.M.

    1988-01-01

    Corrosion is a major service life limiting mechanism for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). For the BWR, stress corrosion cracking of piping has been the major source of concern where extensive research has led to a number of qualified remedies and currently > 90% of susceptible welds have been mitigated or replaced. Stress corrosion cracking of reactor internals due to the interaction of irradiation, as discussed elsewhere in this conference, is also a possible life limiting phenomenon. This paper focusses on two corrosion phenomena in the BWR which have only recently been identified as impacting the universal goal of BWR life extension: the general corrosion of containment structures and the erosion-corrosion of carbon steel piping

  8. Dynamic behaviour of bubbles of water vapour at a temperature lower than the boiling temperature

    International Nuclear Information System (INIS)

    Jansen, Franz

    1966-01-01

    This research thesis reports the study of the theoretical movement of the wall of vapour water bubbles in a sub-saturated boiling regime, i.e. with an average water temperature lower than the boiling temperature. While assuming that bubbles have an initial translational speed at the beginning of their condensation, the author shows that their shrinkage should result in an accelerated displacement in a direction normal to the wall and inward the liquid. Layers of hot water initially close to the wall would therefore be quickly transported towards cold water areas. Experiments allowed, in some cases, the acceleration of bubbles during their condensation to be noticed: for low sub-saturations in still water and for high sub-saturations in water in forced convection, even though, in this last case, the determination of accelerations is more delicate [fr

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

    International Nuclear Information System (INIS)

    Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

    2012-09-01

    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)

  10. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

  11. Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Lofaro, R.J.; Aggarwal, S.

    1992-01-01

    The effects of aging on Residual Heat Removal systems in Boiling Water Reactors have been studied as part of the Nuclear Plant Aging Research Program. The aging phenomena has been characterized by analyzing operating experience from various national data bases. In addition, actual plant data was obtained to supplement and validate the data base findings

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

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1977-01-01

    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

  13. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

  14. Electrochemical potential measurements in boiling water reactors; relation to water chemistry and stress corrosion

    International Nuclear Information System (INIS)

    Indig, M.E.; Cowan, R.L.

    1981-01-01

    Electrochemical potential measurements were performed in operating boiling water reactors to determine the range of corrosion potentials that exist from cold standby to full power operation and the relationship of these measurements to reactor water chemistry. Once the corrosion potentials were known, experiments were performed in the laboratory under electrochemical control to determine potentials and equivalent dissolved oxygen concentrations where intergranular stress corrosion cracking (IGSCC) would and would not occur on welded Type-304 stainless steel. At 274 0 C, cracking occurred at potentials that were equivalent to dissolved oxygen concentration > 40 to 50 ppb. With decreasing temperature, IGSCC became more difficult and only severely sensitized stainless steel would crack. Recent in-reactor experiments combined with the previous laboratory data, have shown that injection of small concentrations of hydrogen during reactor operation can cause a significant decrease in corrosion potential which should cause immunity to IGSCC. (author)

  15. Liquid-solid contact measurements using a surface thermocouple temperature probe in atmospheric pool boiling water

    International Nuclear Information System (INIS)

    Lee, L.Y.W.; Chen, J.C.; Nelson, R.A.

    1984-01-01

    Objective was to apply the technique of using a microthermocouple flush-mounted at the boiling surface for the measurement of the local-surface-temperature history in film and transition boiling on high temperature surfaces. From this measurement direct liquid-solid contact in film and transition boiling regimes was observed. In pool boiling of saturated, distilled, deionized water on an aluminum-coated copper surface, the time-averaged, local-liquid-contact fraction increased with decreasing surface superheat. Average contact duration increased monotonically with decreasing surface superheat, while frequency of liquid contact reached a maximum of approx. 50 contacts/s at a surface superheat of approx. 100 K and decreased gradually to 30 contacts/s near the critical heat flux. The liquid-solid contact duration distribution was dominated by short contacts 4 ms at low surface superheats, passing through a relatively flat contact duration distribution at about 80 0 K. Results of this paper indicate that liquid-solid contacts may be the dominant mechanism for energy transfer in the transition boiling process

  16. Investigation of the minimum film boiling temperature of water during rewetting under forced convective conditions

    International Nuclear Information System (INIS)

    Huang, X.C.; Bartsch, G.; Wang, B.X.

    1992-01-01

    The minimum film boiling temperature of water has been measured on a copper hollow cylinder of 50 mm length with the mass flux rate ranging from 25 to 500 kg/m 2 s and the pressure from 0.1 to 1.0 MPa at subcoolings of 5 to 50 K. Film boiling is established with help of a temperature-controlled system. Rewetting can be initiated by cutting off or very gradually reducing the power supply to the test section. A numerical method for solving the two-dimensional nonlinear inverse heat conduction problem is utilized in the data reduction, taking into account the axial heat conduction. The results are compared with the steady-state maximum transition boiling temperatures measured on the same test section and with the true quench temperatures available in the literature so far. (4 figures, 1 table) (Author)

  17. Output control system in a boiling water atomic power plant

    International Nuclear Information System (INIS)

    Sadakane, Ken-ichiro.

    1975-01-01

    Object: To provide a line in bypass relation with a water heater, a flow rate of said bypass being adjusted to thereby perform quick responsive sub-cool control of a core inlet. Structure: A steam line and a water line are disposed so as to feed water from the reactor core to the water heater via turbine and thence to the core. A line disposed in bypass relation with the water heater arranged in the water line includes a control valve for controlling water passing through the bypass line and a main control for sending a signal to said control valve, said main control receiving loads from the outside, whereby a control signal is transmitted to the control valve, causing water passing through the water heater and water line to the core to be bypassed, a period of time for supplying time to be reduced, and quick response to be enhanced. (Kamimura, M.)

  18. 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)

    Zboray, Robert; Kickhofel, John; Damsohn, Manuel; Prasser, Horst-Michael

    2011-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

    state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This paper presents a methodology to address combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. In order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.

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

    International Nuclear Information System (INIS)

    Brettschuh, Werner

    2006-01-01

    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

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

    International Nuclear Information System (INIS)

    Brettschuh, W.

    2006-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This paper presents a methodology to address combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. In order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.

  3. 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

    ... DEPARTMENT OF HEALTH AND HUMAN SERVICES Food and Drug Administration [Docket No. FDA-2010-D-0236] Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory; Availability AGENCY: Food and Drug Administration, HHS. ACTION: Notice. SUMMARY: The Food and Drug...

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

    International Nuclear Information System (INIS)

    Lanthen, Jonas

    2006-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

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

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

    International Nuclear Information System (INIS)

    Ott, L.J.

    1997-01-01

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

  7. 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)

  8. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

    2008-01-01

    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)

  9. Specifications of the BWR simulator for HAMMLAB 2000

    International Nuclear Information System (INIS)

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

    1998-02-01

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

  10. A BWR 24-month cycle analysis using multicycle techniques

    International Nuclear Information System (INIS)

    Hartley, K.D.

    1993-01-01

    Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

  11. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

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

    International Nuclear Information System (INIS)

    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 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

  13. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  14. Corrosion resistance improvement of ferritic steels through hydrogen additions to the BWR coolant

    International Nuclear Information System (INIS)

    Gordon, B.M.; Jewett, C.W.; Pickett, A.E.; Indig, M.E.

    1984-01-01

    Motivated by the success of oxygen suppression for mitigation of intergranular stress corrosion cracking (IGSCC) in weld sensitized austenitic materials used in Boiling Water Reactors (BWRs), oxygen suppression, through hydrogen additions to the feedwater was investigated to determine its affect on the corrosion resistance of ferritic and martensitic BWR structural materials. The results of these investigations are presented in this paper, where particular emphasis is placed on the corrosion performance of BWR pressure vessel low alloy steels, carbon steel piping materials and martensitic pump materials. It is important to note that the corrosion resistance of these materials in the BWR environment is excellent. Consequently this investigation was also motivated to determine whether there were any detrimental effects of hydrogen additions, as well as to identify any additional margin in ferritic/martensitic materials corrosion performance

  15. Tritium in liquid phase in a BWR-5 like Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Vargas A, A.; Cardenas J, J.

    2011-11-01

    In boiling water reactors (BWR), the tritium (H 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)

  16. 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.

  17. Feedwater processing method in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izumitani, M; Tanno, K

    1976-09-06

    The purpose of the invention is to decrease a quantity of corrosion products moving from the feedwater system to the core. Water formed into vapor after heated in a reactor is fed to the turbine through a main steam line to drive a generator to return it to liquid-state water in a condenser. The water is then again cycled into the reactor via the condensate pump, desalting unit, low pressure feedwater heater, medium pressure feedwater heater, and high pressure feedwater heater. The reactor water is recycled by a recycling pump. At this time, the reactor water recycled by the recycling pump is partially poured into a middle point between the desalting unit and the low pressure feedwater heater through a reducing valve or the like. With the structure described above, the quantity of the corrosion products from the feedwater system may be decreased by the function of a large quantity of active oxygen contained in the reactor water.

  18. Correlation between the solubility of aromatic hydrocarbons in water and micellar solutions, with their normal boiling points

    International Nuclear Information System (INIS)

    Almgren, M.; Grieser, F.; Powell, J.R.; Thomas, J.K.

    1979-01-01

    A linear correlation between the logarithm of the solubility in water of aromatic hydrocarbons and their normal boiling points is shown. Similarly, the logarithm of the distribution ratio of aromatic hydrocarbons in aqueous micellar solution is shown to be linearly related to the boiling points of the hydrocarbons. 2 figures, 2 tables

  19. Early hydrogen water chemistry project review, improvement opportunities and conceptural design options at Exelon boiling water reactors

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Odell, Andrew D.; Giannelli, Joseph F.

    2012-09-01

    Intergranular Stress Corrosion Cracking (IGSCC) and its impacts have been a major concern to the BWR fleet since the mid-70's. Several alternative strategies have been employed to reduce the negative impacts, however, the newest being Early Hydrogen Water Chemistry (EHWC). The Electric Power Research Institute (EPRI) and the BWRVIP (Vessel Internals Project) has strongly supported the development of EHWC, including laboratory testing and a demonstration program that was performed at Peach Bottom Atomic Power Station in October 2011. This paper will review the impacts of a 'Special Test Program' on a BWR plant including: Project management findings; technical reviews and documents required to support such a demonstration program; temporary equipment design, installation and testing; keeping the demonstration progressing along with the plant return from a refuel outage; and lessons learned that can be applied to EHWC implementation during future start-ups. Details will be compared between various Exelon BWRs in support of conceptual designs for EHWC systems and operation. Some comparisons on operational impacts will be provided between various types of BWR plants with differing 'Balance of Plant' designs. (authors)

  20. PWR and BWR light water reactor systems in the USA and their fuel cycle

    International Nuclear Information System (INIS)

    Crawford, W.D.

    1977-01-01

    Light water reactor operating experience in the USA can be considered to date from the choice of the PWR for use in the naval reactor programme and the subsequent construction and operation of the nuclear power plant at Shippingport in 1957. The development of the BWR in 1954 and its selection for the plant at Dresden in 1959 established this concept as the other major reactor type in the US nuclear power programme. The subsequent growth profile is presented. A significant operating record has been accumulated concerning the availability of each of these reactor types. In addition, the use and performance of BWRs and PWRs in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to ensure effective safeguards at nuclear power installations; current measures are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. Both fuel cycles are examined in terms of: fuel burnup experience and prospects for improvement; natural uranium resources; enrichment capacity; reprocessing and recycle; and the interrelationships among the latter three factors. High-level waste management currently involving on-site storage of spent fuel is discussed in terms of available capacity and plans for expansion. The US electric utility industry viewpoint regarding an ultimate programme for waste management is outlined. Finally, the current economics and future cost trends of nuclear power plants are evaluated. (author)

  1. Development of BWR [boiling water reactor] and PWR [pressurized water reactor] event descriptions for nuclear facility simulator training

    International Nuclear Information System (INIS)

    Carter, R.J.; Bovell, C.R.

    1987-01-01

    A number of tools that can aid nuclear facility training developers in designing realistic simulator scenarios have been developed. This paper describes each of the tools, i.e., event lists, events-by-competencies matrices, and event descriptions, and illustrates how the tools can be used to construct scenarios

  2. A study on the feasibility of minor actinides in BWR

    International Nuclear Information System (INIS)

    Abdul Waris; Budiono

    2008-01-01

    Preliminary study on the feasibility of actinides minor (MA) recycling without mixing them with plutonium in boiling water reactor (BWR) has been carried out. The results show that increasing of fissile MA content in mixed oxide fuel (MOX) and/or reducing void fraction can enlarge the effective multiplication factor at the beginning of cycle, but the reactor still can not obtain its criticality condition. Furthermore, dropping the void fraction results in higher reactivity swing and therefore plummeting the safety factor of the reactor. (author)

  3. Thermal analysis and design of a passive reflux condenser for the simplified boiling water reactor

    International Nuclear Information System (INIS)

    Bijlani, C.; Patti, F.; Prasad, V.

    1993-01-01

    At present, the advanced light water reactors (ALWRS) in the United States are being designed to remove reactor decay heat for a period of 72 h following a postulated loss-of-coolant accident (LOCA). The water in the pools external to the containment is evaporated or boiled off to remove the decay heat. It is presumed that the water in the pools can be replenished within 72 h through operator actions or outside assistance. Some countries in Europe require that the plant be designed to remove the reactor decay heat for a much longer duration than 72 h without external assistance. This paper presents an analysis and design of a passive heat exchanger called a reflux condenser (RC), which was considered for an ALWR-the 600-MW(electric) simplified boiling water reactor. The RC is required to condense the steam formed when the water in the pool in which the passive containment cooling system (PCCS) is immersed boils following a LOCA. The RCs are nuclear non-safety related. This paper presents steady-state performance of an RC at various outdoor air dry-bulb temperatures under still air conditions

  4. Detailed pressure drop measurements in single-and two-phase adiabatic air-water turbulent flows in realistic BWR fuel assembly geometry with spacer grids

    International Nuclear Information System (INIS)

    Caraghiaur, Diana; Frid, Wiktor; Tillmark, Nils

    2004-01-01

    In recent years, advance numerical simulation tools based on CFD methods have been increasingly used in various multi-phase flow applications. One of these is two-phase flow in fuel assemblies of Boiling Water Reactors. The important and often missing aspect of this development is validation of CFD codes against proper experimental data. The purpose of the current paper is to present detailed pressure measurements over a spacer grid in low pressure adiabatic single- and bubbly two-phase flow, which will be used to further develop a CFD code for BWR fuel bundle analysis. The experiments have been carried out in a n asymmetric 24-rod sub-bundle, representing one quarter of a Westinghouse SVEA-96 nuclear reactor fuel assembly. Single-phase flow measurements have been performed at superficial velocities between 0.90-4.50 m/s and in the two-phase flow, which was simulated by air-water mixture, measurements have been performed at void fractions ranging from 4 to 12% and liquid superficial velocity of 4.50 m/s. In order to increase the number of measuring points, five pressure taps were drilled in one of the rods, which was easily moved vertically by a traversing system, covering most of the points in axial direction. Any of the rods in the bundle could be substitute by the pressure sensing rod and the measurements were made for five pressure taps facing-angles. A detailed pressure distribution comparison between single- and two-phase flows for different sub-channel positions and different flow conditions was performed over one of the spacers. In addition, single-phase pressure drop measurements in the upper part of the test section comprising two spacer grids have been carried out. (author)

  5. Identification of the two-phase flow in the upper part of a boiling water reactor core using reactor noise analysis

    International Nuclear Information System (INIS)

    Miteff, L.

    1983-08-01

    The starting point of this work were neutron flux correlation measurements in the core of the Muehleberg boiling water reactor. During these measurements a new effect was observed i.e. that the cross power spectral density (CPSD) phases could be approximated by two straight lines of different slopes. The two linear domains of the CPSD-phases are separated along the frequency axis by a transition interval. It was supposed that such CPSD-phases could account for the propagation in the axial direction of two different thermohydraulic perturbations in the upper part of the Muehleberg BWR-core. It was also assumed that the second linear domain of these CPSD-phases could be related to a characteristic property of an annular flow regime of the steam-water two-phase flows in the bundles around the neutron detectors. The author attempts to give an explanation for the existence of a second transport phenomenon. This was achieved by in-core and out-core correlation measurements as well as by theoretical work. The out-core measurements were performed on an water-air simulation loop by use of laser beams. (Auth.)

  6. Numerical Simulation on Subcooled Boiling Heat Transfer Characteristics of Water-Cooled W/Cu Divertors

    Science.gov (United States)

    Han, Le; Chang, Haiping; Zhang, Jingyang; Xu, Tiejun

    2015-04-01

    In order to realize safe and stable operation of a water-cooled W/Cu divertor under high heating condition, the exact knowledge of its subcooled boiling heat transfer characteristics under different design parameters is crucial. In this paper, subcooled boiling heat transfer in a water-cooled W/Cu divertor was numerically investigated based on computational fluid dynamic (CFD). The boiling heat transfer was simulated based on the Euler homogeneous phase model, and local differences of liquid physical properties were considered under one-sided high heating conditions. The calculated wall temperature was in good agreement with experimental results, with the maximum error of 5% only. On this basis, the void fraction distribution, flow field and heat transfer coefficient (HTC) distribution were obtained. The effects of heat flux, inlet velocity and inlet temperature on temperature distribution and pressure drop of a water-cooled W/Cu divertor were also investigated. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor. supported by the National Magnetic Confinement Fusion Science Program of China (No. 2010GB104005), Funding of Jiangsu Innovation Program for Graduate Education (CXLX12_0170), the Fundamental Research Funds for the Central Universities of China

  7. THYDE-B1/MOD2: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Muramatsu, Ken; Kukita, Yutaka; Tasaka, Kanji

    1988-04-01

    THYDE-B1/MOD2 is a fast-running best estimate (BE) computer code to analyze thermal-hydraulic behaviors of the reactor cooling system of a boiling water reactor (BWR), mainly, during a small-break loss-of-coolant accident (SBLOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions consist of subcooled liquid, saturated mixture and saturated steam regions from the volume bottom. The regions are separated by two horizontal moving boundaries which are tracked by mass and energy balances for each region. With this three region node model, the interior of the pressure vessel can be represented by only two volumes: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous node model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SBLOCAs in which the thermal-hydraulic behavior is relatively slow and gravity controlled. The code has been improved and modified from the last version of the code, THYDE-B1/MOD1, especially in the phase separation model which is used in the mixture level calculation in the three region node model. Then, a good predictability of the code has been indicated through the comparison of calculated results with various SBLOCA test data including ROSA-III of JAERI and FIST of the General Electric Co. This report presents the code modifications and input data requirements of the THYDE-B1/MOD2 code. (author)

  8. Advanced Best-Estimate Methodologies for Thermal-Hydraulics Stability Analyses with TRACG code and Improvements on Operating Boiling Water Reactors

    International Nuclear Information System (INIS)

    Vedovi, J.; Trueba, M.; Ibarra, L; Espino, M.; Hoang, H.

    2016-01-01

    In recent years GE Hitachi has introduced two advanced methodologies to address the thermal-hydraulics instabilities in Boiling Water Reactors (BWRs); the “Detect and Suppress Solution - Confirmation Density (DSS-CD)” and the “GEH Simplified Stability Solution (GS3).” These two methodologies are based on Best-Estimate Plus Uncertainty (BEPU) analyses and provide significant improvement on safety, plant maneuvering and fuel economics with respect to existing solutions. DSS-CD and GS3 solutions have been recently approved by the United States Nuclear Regulatory Commission. This paper describes the main characteristics of these two stability methodologies and shares the experience of their recent implementation in operating BWRs. The BEPU approach provided a much deeper understanding of the parameters affecting instabilities in operating BWRs and allowed for better calculation of plant setpoints by improving plant manoeuvring restrictions and reducing manual operator actions. DSS-CD and GS3 methodologies are both based on safety analyses performed with the best-estimate system code TRACG. The assessment of uncertainty is performed following the Code Scaling, Applicability and Uncertainty (CSAU) methodology documented in NUREG/CR-5249. The two solutions have been already implemented in a combined 18 BWR units with 7 more units in the process of transitioning. The main results demonstrate a significant decrease (>0.1) in the stability based Operating Limit Minimum Critical Power Ratio (OLMCPR), which possibly results in significant fuel savings and the increase in allowable stability plant setpoints that address instability events such as the one occurred at the Fermi 2 plant in 2015 and can help prevent unnecessary Scrams. The paper also describes the advantages of reduced plant manoeuvring as a result to transitioning to these solutions; in particular the history of a BWR/6 transition to DSS-CD is discussed.

  9. Metallurgical factors that contribute to cracking in BWR piping

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

  10. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

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

    2010-01-01

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

  11. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

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

    International Nuclear Information System (INIS)

    Boing, L.E.

    1989-12-01

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

  13. Power distribution monitoring system in the boiling water cooled reactor core

    International Nuclear Information System (INIS)

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

    1987-01-01

    Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

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

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.

    1989-12-01

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

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

    International Nuclear Information System (INIS)

    Gavilian-Moreno, Carlos; Espinosa-Paredes, Gilberto

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  17. Using Largest Lyapunov Exponent to Confirm the Intrinsic Stability of Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Carlos J. Gavilán-Moreno

    2016-04-01

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

  18. Use of adaptive diffusion theory based monitors in optimizing boiling water reactor core designs

    International Nuclear Information System (INIS)

    Congdon, S.P.; Martin, C.L.; Crowther, R.L.

    1988-01-01

    Three-dimensional coarse mesh models are routinely used to predict the performance of boiling water reactors. In the adaptive monitory model, the three-dimensional solutions are permanently adapted to incore probe data. The corrections resulting from the adaptive process lead to reliable predictions of future reactor states. The corrections can also be carried forward to future operating cycles. This can shorten the time required to introduce an validate new design and operating strategy improvements. (orig.) [de

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

    International Nuclear Information System (INIS)

    Kamei, Kazuhiro; Tanoue, Tetsuharu; Kataoka, Kazuyoshi; Jimbo, Masakazu

    2011-01-01

    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)

  20. Determining the void fraction in draught sections of a boiling water cooled reactor

    International Nuclear Information System (INIS)

    Fedulin, V.N.; Barolomej, G.G.; Solodkij, V.A.; Shmelev, V.E.

    1987-01-01

    Consideration is being given to the problem of improving methods for calculation of the void fraction in large channels of cooling system of the boiling water cooled reactor during two-phase unsteady flow. Investigation of the structure of two-phase flow was conducted in draught section of the VK-50 reactor (diameter D=2 m, height H=3). The method for calculation of the void fraction in channels with H/D ratio close to 1 is suggested

  1. Evaluation of internal flooding in a BWR

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  2. Saturated flow boiling heat transfer in water-heated vertical annulus

    International Nuclear Information System (INIS)

    Sun Licheng; Yan Changqi; Sun Zhonning

    2005-01-01

    This paper describes the saturated flow boiling heat transfer characteristics of water at 1 atm and low velocities in water-heated vertical annuli with equivalent diameters of 10 mm and 6 mm. Test section is consisted of two concentric circular tubes outer of which is made of quartz, so the whole test courses can be visualized. There are three main flow patterns of bubble flow, churn flow and churn-annular flow in the annuli, most important of which is churn flow. Flooding is the mechanism of churn flow and churn can enhance the heat transport between steam and water; Among the three factors of mass flux, inlet subcooling and annulus width, the last one has great effect on heat transport, moderately decreasing the annulus width can enhance the heat transfer; Combined annular flow model with theory of flooding and turbulent Prandtl Number, the numerical value of heat flux is given, the shape of test boiling curve and that of calculated by model is very alike, but there is large discrepancy between test data and calculated results, the most possible reason is that some parameters given by fluid flooding model are based on experimental data of common circular tubes, but not of annuli. Doing more research on flooding in annulus, particularly narrow annulus, is necessary for calculating the saturated boiling in annulus. (authors)

  3. Forced convective boiling of water inside helically coiled tube. Characteristics of oscillation of dryout point

    International Nuclear Information System (INIS)

    Nagai, Niro; Sugiyama, Kenta; Takeuchi, Masanori; Yoshikawa, Shinji; Yamamoto, Fujio

    2006-01-01

    The helically coiled tube of heat exchanger is used for the evaporator of prototype fast breeder reactor 'Monju'. This paper aims at the grasp of two-phase flow phenomena of forced convective boiling of water inside helical coiled tube, especially focusing on oscillation phenomena of dryout point. A glass-made helically coiled tube was used to observe the inside water boiling behavior flowing upward, which was heated by high temperature oil outside the tube. This oil was also circulated through a glass made tank to provide the heat source for water evaporation. The criterion for oscillation of dryout point was found to be a function of inlet liquid velocity and hot oil temperature. The observation results suggest the mechanism of dryout point oscillation mainly consists of intensive nucleate boiling near the dryout point and evaporation of thin liquid film flowing along the helical tube. In addition, the oscillation characteristics were experimentally confirmed. As inlet liquid velocity increases, oscillation amplitude also increases but oscillation cycle does not change so much. As hot oil temperature increases, oscillation amplitude and cycle gradually decreases. (author)

  4. Fuel Element Experience at the Halden Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aas, S. [OECD Halden Reactor Project, Halden (Norway); Videm, K.; Hanevik, A. [Institutt for Atomenergi, Kjeller (Norway)

    1968-04-15

    The penalty for neutron absorbing materials is higher for a reactor moderated with heavy water than one with light water. As Zircaloy and enriched uranium were not readily available in 1954 when the design of the first fuel charge for HBWR was frozen, fuel elements of natural uranium metal clad in a specially developed aluminium alloy (A 1 0.3% Fe, 0.03% Si) were used. The temperature was limited to 150 Degree-Sign C and with this limitation the general behaviour of the elements was good. In I960, in another effort to maintain a good neutron economy, a couple of elements with as thin cladding as 0.25 mm A1S1 316, stainless steel with an unsegmented length of 2 m supported by wire grid spacers were tested. These elements with 1.5% enriched UO{sub 2} behaved satisfactorily at 150'C. Elements of a rather similar construction failed due to stress corrosion during the later operation at 230 'C. The reason for the different behaviour is probably the higher stresses in the cladding, due to the increased pressure, possibly combined with a short period with a high chloride content in the heavy water. The second fuel core with 1.5% enriched UO{sub 2} clad in Zircaloy-2 was installed in order to permit an increase in temperature to 230 Degree-Sign C and in power from 5 to 20 MW(th). The maximum burnup obtained is 11000 MWd/t and the maximum heat rating 375 W/cm with no fracture failure and practically no change in appearance according to the post-irradiation examination. One element was deliberately taken to burn-out conditions by throttling the water flow. After a series of burn-outs, the element finally failed because of over-temperature. The successful use of aluminium cladding at 150 Degree-Sign C mitiated an effort for making aluminium alloys suitable for normal power reactor operation. Promising properties were found for an alloy (designated IFA 3 aluminium) with A1 10% Si, 1% Ni, 1% Mg, 0.3% Fe + Ti. Despite increase in corrosion rate under heat transfer conditions

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

    International Nuclear Information System (INIS)

    Freydier, P.; Chen, O.; Olive, J.; Simonin, O.

    1991-04-01

    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

  6. RELAP 4/MOD 6 boiling water nodalization study

    International Nuclear Information System (INIS)

    Sonneck, G.; Pfau, H.

    1985-09-01

    The risk of nuclear steam supply systems is dominated by the core melt accidents. The first step to a realistic assessment of these sequences is the successful prediction of a loss of coolant event in a test loop. One of the codes for that is RELAP 4/MOD 6 and one of the important options in this code is the nodalization. The base of this work is the test LOCA No. 1 FIX II in Studsvik (Sweden) which also served as the OECD International Standard Problem 15. This report discusses the influence of different nodalizations, of different distributions of pressure, water and structural heat as well as of different bubble rise options, break flow coefficients, and heat transfer time steps. The most important result is that a simple RELAP 4/MOD6 model with less than 10 volumes is able to predict an experiment as LOCA No. 1 in FIX II successfully using only a fraction of the usual computing time. (Author)

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

    International Nuclear Information System (INIS)

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-01-01

    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

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

    International Nuclear Information System (INIS)

    Elliott, H.H.; Holloway, J.H.; Abbott, D.G.

    1979-01-01

    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

  9. Evaluation of leachable behavior from ion exchange resins effects of organic impurities on BWR water chemistry. 1

    International Nuclear Information System (INIS)

    Igarashi, Hiroo; Nishimura, Yusaku; Ohsumi, Katsumi; Uchida, Shunsuke; Matsui, Tsuneo

    1999-01-01

    The elution rate of leachables from ion exchange resin, which is used in condensate demineralizers and is one of several major sources of organic compounds in BWR cooling water, was measured. Properties of the leachables and elution rate depended on the kind of ion exchange resin and the years of use. The organic compounds elution rate of cation exchange resin was constant for 5 years and the molecular weight of these leachables was low. After 5 years, the elution rate increased and leachables consisted of organic compounds of high molecular weights of several thousand. The elution rate of anion exchange resin decreased yearly. The difference in the elution behavior was attributed to a dependence on oxidation degradation promoted by transition metal catalysis. The cation exchange resin included absorbed transition metal, while the anion exchange resin did not. An empirical formula showing the time dependence of the elution rate of organic compounds was derived. The formula was judged to be appropriate based on simulations of an actual BWR plant and comparisons of impurity concentrations with actual reactor water data. (author)

  10. Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.

  11. Thermal Esophageal Injury following Ingestion of Boiling Mushroom Water

    Directory of Open Access Journals (Sweden)

    Allison Prevost

    2017-01-01

    Full Text Available Thermal esophageal and gastric damage from ingestion of hot liquids is poorly studied in pediatrics. Limited case reports exist in the literature. Many cases presented with chest pain, dysphagia, and odynophagia. Variable histologic findings were reported. No definitive management guidelines exist for such injuries. We provide a report of the acute assessment and management of an obvious thermal esophageal injury and contribute to what is known about this presentation. A 16-year-old male presented with odynophagia, dysphagia, and hematemesis following ingestion of “nearly boiling” mushroom water. Ondansetron, pantoprazole, ketorolac, maintenance intravenous fluids, and a clear liquid diet were started. At sixty hours after ingestion, an esophagogastroduodenoscopy (EGD revealed blistering and edema of the soft palate and epiglottis, circumferential erythema of the entire esophagus with an exudate likely to be desquamated mucosa, and linear erythema of the body and fundus of the stomach. An EGD one month after ingestion showed no residual effects from the injury. The pantoprazole was weaned and restrictions to his diet were lifted. To better standardize care in these rare esophageal injuries, the development of a clinical care algorithm may be beneficial to provide clinicians with a guide for management based on outcomes of previously reported cases.

  12. Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Venker, Jeanne

    2015-03-31

    The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO{sub 2}) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO{sub 2} Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO{sub 2} Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO{sub 2} Brayton cycles and to evaluate the introduced

  13. Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

    International Nuclear Information System (INIS)

    Venker, Jeanne

    2015-01-01

    The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO 2 ) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO 2 Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO 2 Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO 2 Brayton cycles and to evaluate the introduced heat removal system

  14. Non-isothermal desorption and nucleate boiling in a water-salt droplet LiBr

    Directory of Open Access Journals (Sweden)

    Misyura Sergey Ya.

    2018-01-01

    Full Text Available Experimental data on desorption and nucleate boiling in a droplet of LiBr-water solution were obtained. An increase in salt concentration in a liquid-layer leads to a considerable decrease in the rate of desorption. The significant decrease in desorption intensity with a rise of initial mass concentration of salt has been observed. Evaporation rate of distillate droplet is constant for a long time period. At nucleate boiling of a water-salt solution of droplet several characteristic regimes occur: heating, nucleate boiling, desorption without bubble formation, formation of the solid, thin crystalline-hydrate film on the upper droplet surface, and formation of the ordered crystalline-hydrate structures during the longer time periods. For the final stage of desorption there is a big difference in desorption rate for initial salt concentration, C0, 11% and 51%. This great difference in the rate of desorption is associated with significantly more thin solution film for C0 = 11% and higher heat flux.

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

    International Nuclear Information System (INIS)

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Sulatsky, A.A.; Khabensky, V.B.; Lopukh, D.B.; Petrov, Y.B.; Pechenkov, A.Y.

    2000-01-01

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the 'Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO 2+x -16% ZrO 2 -15% Fe 2 O 3 -6% Cr 2 O 3 -3% Ni 2 O 3 . The melt surface temperature ranged within 1920-1970 K. (orig.)

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

    International Nuclear Information System (INIS)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.

    1999-01-01

    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%UO 2 - 16%ZrO 2 - 15%Fe 2 O 3 - 6%Cr 2 O 3 -3%Ni 2 O 3 . The melt surface temperature was 1650-1700degC. (author)

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

    International Nuclear Information System (INIS)

    Osmachkin, V.S.; Sokolov, I.N.

    1998-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, Hans-Peter

    1977-08-15

    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

  2. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

    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 140 La 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

  3. An investigation of transition boiling mechanisms of subcooled water under forced convective conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kwang-Won, Lee; Sang-Yong, Lee

    1995-09-01

    A mechanistic model for forced convective transition boiling has been developed to investigate transition boiling mechanisms and to predict transition boiling heat flux realistically. This model is based on a postulated multi-stage boiling process occurring during the passage time of the elongated vapor blanket specified at a critical heat flux (CHF) condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling characterized by the frequent touches of the interface and the heated wall. The total heat transfer rates after the DNB is weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. The parametric effects of pressure, mass flux, inlet subcooling on the transition boiling heat transfer are also investigated. From these comparisons, it can be seen that this model can identify the crucial mechanisms of forced convective transition boiling, and that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are well predicted at low qualities/high pressures near 10 bar. In future, this model will be improved in the unstable film boiling stage and generalized for high quality and low pressure situations.

  4. Enhanced Natural Convection in a Metal Layer Cooled by Boiling Water

    International Nuclear Information System (INIS)

    Cho, Jae-Seon; Suh, Kune Y.; Chung, Chang-Hyun; Park, Rae-Joon; Kim, Sang-Baik

    2004-01-01

    An experimental study is performed to investigate the natural convection heat transfer characteristics and the solidification of the molten metal pool concurrently with forced convective boiling of the overlying coolant to simulate a severe accident in a nuclear power plant. The relationship between the Nusselt number (Nu) and the Rayleigh number (Ra) in the molten metal pool region is determined and compared with the correlations in the literature and experimental data with subcooled water. Given the same Ra condition, the present experimental results for Nu of the liquid metal pool with coolant boiling are found to be higher than those predicted by the existing correlations or measured from the experiment with subcooled boiling. To quantify the observed effect of the external cooling on the natural convection heat transfer rate from the molten pool, it is proposed to include an additional dimensionless group characterizing the temperature gradients in the molten pool and in the external coolant region. Starting from the Globe and Dropkin correlation, engineering correlations are developed for the enhancement of heat transfer in the molten metal pool when cooled by an overlying coolant. The new correlations for predicting natural convection heat transfer are applicable to low-Prandtl-number (Pr) materials that are heated from below and solidified by the external coolant above. Results from this study may be used to modify the current model in severe accident analysis codes

  5. Passive cooling of condensate chambers as retrofitting measure in boiling water reactors; Passive Kuehlung der Kondensationskammern in Siedewasserreaktoren als Nachruestmassnahme

    Energy Technology Data Exchange (ETDEWEB)

    Freis, Daniel; Nachtrodt, Frederik; Sporn, Michael; Tietsch, Wolfgang; Sassen, Felix [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2012-11-01

    Westinghouse Electric Germany GmbH has developed a concept for passive cooling of condensate chambers of BWR-type reactors. Due to its compactness the system is feasible as retrofitting measure. The passive condensate chamber cooling system is based on a cooling module with ascending and down pipe that are connected with the evaporation condenser to form a cooling circuit. Based on the consequent use of high-effective heat transport mechanisms, as boiling, condensation without non-condensable gases and mass transport a high cooling performance and compact construction is possible. The system is completely passive and completely diverse to existing active cooling systems. In the frame of a true-scale experiment the significant cooling performance was demonstrated. RELAP5 calculations confirmed the functionality of the cooling module.

  6. Comparison of finite element and influence function methods for three-dimensional elastic analysis of boiling water reactor feedwater nozzle cracks

    International Nuclear Information System (INIS)

    Besuner, P.M.; Caughey, W.R.

    1976-11-01

    The paper compares the finite element (FE) and influence function (IF) methods for a three-dimensional elastic analysis of postulated circular-shaped surface cracks in the feedwater nozzle of a typical boiling water reactor (BWR). The FE method is incorporated in a direct manner. The nozzle and crack geometry and the complex loading are all included in the model which simulates the structural crack problem. The IF method is used to compute stress intensity factors only when the uncracked stress field (that is, the stress in the uncracked solid at the locus of the crack to be eventually considered) has been computed previously. The IF method evaluates correctly the disturbance of this uncracked stress field caused by the crack by utilizing a method of elastic superposition. Both the IF and FE methods are described in detail in the paper and are applied to several test cases chosen for their similarity to the nozzle crack problem and for the availability of an accurate published result obtained from some recognized third method of solution. Results are given which summarize both the accuracy and the direct computer costs of the two methods

  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)

    Konzek, G.J.; Smith, R.I.

    1990-12-01

    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. Comparison of depletion results for a boiling water reactor fuel element with CASMO and SCALE 6.1 (TRITON/NEWT)

    International Nuclear Information System (INIS)

    Mesado, C.; Morera, D.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, Amparo; Melara, Jose

    2013-01-01

    In this work, the results of depletion calculations with CASMO and SCALE 6.1 (TRITON) are compared. To achieve it, a region of a Boiling Water Reactor (BWR) fuel element is modeled, using both codes. To take into account different operating conditions, the simulations are repeated with different void fraction, ranging from null void fraction to a void fraction closed to one. Special care was used to keep in mind that the homogenization of the materials and the two group approach was the same in both codes. Additionally, KENO-VI and MCDANCOFF modules are used. The k-effective calculated by KENO-VI is used to ensure that the starting point was correct and MCDANCOFF module is used to calculate the Dancoff factors in order to improve the model accuracy. To validate the whole process, a comparison of k eff , and cross-sections collapsed and homogenized is shown. The results show a very good agreement, with an average error around the 1.75%. Furthermore, an automatic process for translating CASMO data to SCALE input decks was developed. The reason for the translation is the fact that SCALE's TRITON module is a new code very powerful and continuously being developed. Thus, great advantage can be taken from the current and future SCALE features. This is hoped to produce more realistic models, and hence, increase the accuracy of neutronic libraries. (author)

  9. Benchmarking lattice physics data and methods for boiling water reactor analysis

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Edenius, M.; Harris, D.R.; Hebert, M.J.; Kapitz, D.M.; Pilat, E.E.; VerPlanck, D.M.

    1983-01-01

    The objective of the work reported was to verify the adequacy of lattice physics modeling for the analysis of the Vermont Yankee BWR using a multigroup, two-dimensional transport theory code. The BWR lattice physics methods have been benchmarked against reactor physics experiments, higher order calculations, and actual operating data

  10. NATBWR: a steady-state model for natural circulation in boiling-water reactors

    International Nuclear Information System (INIS)

    Healzer, J.M.; Abdollahian, D.

    1983-02-01

    This report documents the NATBWR steady-state BWR natural-circulation model and activities under EPRI Project RP1561-1 to gather data and predict the natural-circulation operation of the BWR. The report is organized into two parts, with the first part describing the NATBWR model and applications of the model to available BWR natural-circulation data and the second part providing user and programming information about the model. Five different operating BWR's were selected to demonstrate the application of the NATBWR model, one of each type from BWR/1 through BWR/4. For each operating plant, the available natural circulation data has been compared to model predictions. In addition to the data predictions, the behavior of the BWR system at reduced inventory, where the system is isolated and scrammed, and cooling provided by natural circulation has been analyzed. Finally, included as an appendix to Part 1 of this report is a discussion of the stability of the BWR system at natural-circulation conditions

  11. 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)

  12. Halden Boiling Water Reactor. Plant Performance and Heavy-Water Management

    Energy Technology Data Exchange (ETDEWEB)

    Aas, S.; Jamne, E.; Wullum, T.; Fjellestad, K. [Institutt for Atomenergi, OECD Halden Reactor Project, Halden (Norway)

    1968-04-15

    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/cm{sup 2}. 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 UO{sub 2} : the most highly exposed elements have reached 10000 MWd/t UO{sub 2}. 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 {mu}C/cm{sup 3}. 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

  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. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Science.gov (United States)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  15. Enhancing load-following and/or spectral shift capability in single-sparger natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1992-01-01

    This patent describes a method for obtaining load-following capability in a coiling water reactor (BWR) wherein housed within a reactor pressure vessel (RPV) is a nuclear core disposed within a shroud having a shroud head and which with the RPV defines an annulus region disposed beneath the nuclear core, an upper steam dome connected to a steam outlet in the RPV, a core upper plenum formed within the shroud head and disposed atop the nuclear core, a chimney mounted atop the shroud head and in fluid communication with the core upper plenum and with a steam separator having a skirt which is in fluid communication with the steam dome, the region outside of the chimney defining a downcomer region, there being a water level established therein under normal operation of the BWR, and the RPV containing a feedwater inlet. It comprises: disposing a single sparger connected to the feedwater inlet above the steam separator skirt bottom about the interior circumference of the RPV at an elevation at approximately the water level established during normal operation of the BWR; and adjusting the feedwater flow through the inlet and into the sparger to vary the water level to be above, at or below the elevational location of the sparger in response to load-following need

  16. 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.

  17. Comparison of boiling and chlorination on the quality of stored drinking water and childhood diarrhoea in Indonesian households.

    Science.gov (United States)

    Fagerli, K; Trivedi, K K; Sodha, S V; Blanton, E; Ati, A; Nguyen, T; Delea, K C; Ainslie, R; Figueroa, M E; Kim, S; Quick, R

    2017-11-01

    We compared the impact of a commercial chlorination product (brand name Air RahMat) in stored drinking water to traditional boiling practices in Indonesia. We conducted a baseline survey of all households with children 1000 MPN/100 ml (RR 1·86, 95% CI 1·09-3·19) in stored water than in households without detectable E. coli. Although results suggested that Air RahMat water treatment was associated with lower E. coli contamination and diarrhoeal rates among children water treatment by boiling, Air RahMat use remained low.

  18. Operational experiences with on line BWR condenser tube leak verification

    International Nuclear Information System (INIS)

    Bryant, R.A.; Duvall, W.E.; Kirkley, W.B.; Zavadoski, R.W.

    1988-01-01

    Verifying condenser tube leaks at a boiling water reactor is, at best, a difficult task carried out in hot steamy water boxes with concurrent radiation exposure. For small apparent leaks with slight chemical changes there is always uncertainty of whether the problem is a condenser tube leak or a feedback from radwaste. Most conventional methods (e.g soap tests, Saran wrap suction, and helium tests) usually involve a load reduction to isolate the water boxes one at a time and hours of drain down on each box. The sensitivity of the most sensitive test (helium) is of the order of 7500 l per day per box. Sulfur hexafluoride has been successfully used at a BWR to identify one leaking water box out of four while the unit was at 100 % power. The actual tubes leakig in the water box were identified by injecting helium during drain down of the box and subsequent manifold testing. Additional tests with sulfur hexafluoride on the second BWR unit indicated tight water boxes to within the sensitivity of the measurement, i.e. less than 19 l per day for all four boxes. Problems encountered in both tests included sulfur hexafluoride carry over from the plume of the cooling towers and off gas considerations. In brief sulfur hexafluoride can be used to quickly identify which particular water box has a condenser tube leak or, just as quickly, establish the integrity of all the water boxes to a level not previously attainable. (author)

  19. A method of simulating voids in experimental studies of boiling water reactors

    International Nuclear Information System (INIS)

    Down, H.J.; Dickie, J.; Fox, W.N.

    1963-11-01

    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/cm 3 may be produced with the introduction of only about 0.4 gm/cm 3 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)

  20. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2016-01-01

    Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  1. Burnout in the boiling of water and freon-113 on tubes with annular fins

    International Nuclear Information System (INIS)

    Rubin, I.R.; Pul'kin, I.N.; Roizen, L.I.

    1986-01-01

    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

  2. TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor

    International Nuclear Information System (INIS)

    Martin, R.P.

    1993-05-01

    Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions

  3. Dynamic analysis of the condensate feedwater system in boiling water reactor plants

    International Nuclear Information System (INIS)

    Tanji, J.; Omori, T.

    1982-01-01

    The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

  4. Electrochemical measurements and modeling predictions in boiling water reactors under various operating conditions

    International Nuclear Information System (INIS)

    Indig, M.E.

    1991-01-01

    One important issue for providing life extension to operating boiling water nuclear reactors (BWRs) is the control of stress corrosion cracking in all sections of the primary coolant circuit. This paper links experimental and theoretical methods that provide understanding and measurements of the critical parameter, the electrochemical potential (ECP), and its application to determining crack growth rate among and within the family of BWRs. Measurement of in-core ECP required the development of a new family of radiation-resistant sensors. With these sensors, ECPs were measured in the core and piping of two operating BWRs. Concurrent crack growth measurements were used to benchmark a crack growth prediction algorithm with measured ECPs

  5. Design and performance of General Electric boiling water reactor main steam line isolation valves

    International Nuclear Information System (INIS)

    Rockwell, D.A.; van Zylstra, E.H.

    1976-08-01

    An extensive test program has been completed by the General Electric Company in cooperation with the Commonwealth Edison Company on the basic design type of large main steam line isolation valves used on General Electric Boiling Water Reactors. Based on a total of 40 tests under simulated accident conditions covering a wide range of mass flows, mixture qualities, and closing times, it was concluded that the commercially available valves of this basic type will close completely and reliably as required. Analytical methods to predict transient effects in the steam line and valve after postulated breaks were refined and confirmed by the test program

  6. An Analysis of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Persson, P.

    1963-06-01

    A method of predicting the burnout conditions for flow of boiling water in vertical round ducts is presented. The analysis predicts that the burnout conditions are independent of the L/d-ratio and the inlet temperature, and that the burnout steam quality decreases with increasing surface heat flux and increasing mass velocity. It was also found that the burnout steam quality at low pressures increases with the pressure and reaches a maximum at approximately 70 kg/cm, and thereafter decreases with a further increase of the pressure. The theoretical result compares very well with experimental data from different sources

  7. An Analysis of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Persson, P

    1963-06-15

    A method of predicting the burnout conditions for flow of boiling water in vertical round ducts is presented. The analysis predicts that the burnout conditions are independent of the L/d-ratio and the inlet temperature, and that the burnout steam quality decreases with increasing surface heat flux and increasing mass velocity. It was also found that the burnout steam quality at low pressures increases with the pressure and reaches a maximum at approximately 70 kg/cm, and thereafter decreases with a further increase of the pressure. The theoretical result compares very well with experimental data from different sources.

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

    International Nuclear Information System (INIS)

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

    1992-02-01

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

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

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1978-01-01

    A 1 / 5 -scale test facility of a pressure-suppression system from a Mark I boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy, dynamic loading data during a hypothetical loss-of-coolant accident. A total of 27 air tests have been completed with an average of 175 transducers recorded for each test. An end-to-end calibration of the total measurement system was run to establish accuracy of the data. The instrumentation verified the analysis of the dynamic loading of the pressure-suppression system

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

    International Nuclear Information System (INIS)

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

    1993-08-01

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

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

    International Nuclear Information System (INIS)

    Lee, Bom Soon.

    1994-01-01

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

  12. Heat transfer with water in forced convection without boiling in small diameter tubes

    International Nuclear Information System (INIS)

    Ricque, Roger; Siboul, Roger

    1969-01-01

    This note presents the measurements performed for the establishment of an empirical heat transfer law for water in forced convection without boiling in small diameter tubes (2 and 4 mm), with high flow velocity and strong heat flux, and for relatively low fluid temperatures. A correlation of experimental points is obtained with a very small maximum dispersion: Nu fl = 0,0092 Re fl 0,88 Pr 0,5 (μ fl /μ p ) 0,14 . A correlation for the fiction coefficient is also presented [fr

  13. Calculation of optimum control rod operation programme for boiling water reactor

    International Nuclear Information System (INIS)

    Fehr, L.

    1978-01-01

    Control rod operation programmes are calculated based on a three dimensional Boiling Water Reactor situation model. The position of the control rods at variosu burn-ups is chosen by an optimisation so that the sum of the square deviations of the load density distribution from an optimum distribution ('Haling' distribution) are minimised. Other conditions are remaining critical and observing the thermal limits for central fuel element melting and critical heat surface loading. As an example, an optimum control rod operation programme for the first cycle in Lengen nuclear power station is calculated and is compared with the programme actually used. (orig.) 891 HP [de

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

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Lisitza, F.D.; Romenkov, A.A.; Tokarev, Y.I.

    1998-01-01

    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)

  15. The BWR vessel and internals project - 2001 and beyond

    International Nuclear Information System (INIS)

    Wagoner, V.; Mulford, T.

    2001-01-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  16. The BWR vessel and internals project - 2001 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

    2001-07-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  17. Sensitivity of BWR shutdown margin tests to local reactivity anomalies

    International Nuclear Information System (INIS)

    Cokinos, D.M.; Carew, J.F.

    1987-01-01

    Successful shutdown margin (SDM) demonstration is a required procedure in the startup of a newly configured boiling water reactor (BWR) core. In its most reactive condition throughout a cycle, a BWR core must be capable of being made subcritical by a specified margin with the highest worth control rod fully withdrawn and all other rods at their fully inserted positions. Two different methods are used to demonstrate SDM: (a) the adjacent-rod test and (b) the in-sequence test. In the adjacent-rod test, the strongest rod is fully withdrawn and an adjacent rod is withdrawn to reach criticality. In the in-sequence test, control rods spread throughout the core are withdrawn in a predetermined sequence of withdrawals. Larger than expected core k/sub eff/ values have been observed during the performance of BWR SDM tests. The purpose of the work summarized in this paper has been to investigated and quantify the sensitivity of both the adjacent-rod and in-sequence SDM tests to local reactivity anomalies. This was accomplished by introducing reactivity perturbations at selected four-bundle cell locations and by evaluating their effect on core reactivity in each of the two tests

  18. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake

    1995-01-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data

  19. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

    1995-03-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

  20. Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

    International Nuclear Information System (INIS)

    Wheeler, J.K.; Pallotta, A.S.

    1992-01-01

    The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

  1. 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.

  2. Development of internal CRD for next generation BWR-endurance and robustness tests of ball-bearing materials in high-pressure and high-temperature water

    International Nuclear Information System (INIS)

    Shoji Goto; Shuichi Ohmori; Michitsugu Mori; Shohei Kawano; Tadashi Narabayashi; Shinichi Ishizato

    2005-01-01

    An internal CRD using a heatproof ceramics insulated coil is under development to be a competitive and higher performance as Next- Generation BWR. In the case of the 1700MWe next generation BWR, adapting the internal CRDs, the reactor pressure vessel is almost equivalent to that of 1356 MWe ABWR. The endurance and robustness tests were examined in order to confirm the durability of the bearing for the internal CRD. The durability of the ball bearing for the internal CRD was performed in the high-pressure and high-temperature reactor water of current BWR conditions. The experimental results confirmed the durability of rotational numbers for the operation length of 60 years. We added the cruds into water to confirm the robustness of the ball bearing. The test results also showed good robustness even in high-density crud conditions, compared with the current BWR. This program is conducted as one of the selected offers for the advertised technical developments of the Institute of Applied Energy founded by METI (Ministry of Economy, Trade and Industry) of Japan. (authors)

  3. Innovative nuclear technologies based on radiation induced surface activation. (5) Development of high performance BWR by the radiation induced surface activation visualization study on the boiling enhancement with irradiation

    International Nuclear Information System (INIS)

    Imai, Yasuyuki; Okamoto, Koji; Madarame, Haruki; Takamasa, Tomoji

    2004-01-01

    Improvement of CHF requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, we confirmed that the surface wettability changed significantly or that highly hydrophilic conditions were achieved, after irradiation of 60 Co gamma ray, by the Radiation Induced Surface Activation (RISA) phenomenon. To delineate the effect of RISA on boiling phenomena, CHF of metal oxides irradiated by gamma rays were investigated. The heating test section made of titanium was 0.5 mm in diameter. Oxidation of the surface was carried out by plasma jetting. The test section was irradiated by 60 Co gamma ray with predetermined radiation intensity and period. A test piece had been hold horizontally on the electrode after 5400 kGy irradiation. Then, the whole CHF test apparatus with test piece was set on the table in the gamma ray irradiation room. The test piece was irradiated in the water at least 30 minutes. A CHF experiment in the pool boiling condition was carried out under atmospheric pressure under irradiation. The results of on-site experiment were compared with that of off-site one. (author)

  4. 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)

    Membrillo G, O. E.; Chavez M, C.

    2012-10-01

    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)

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

    International Nuclear Information System (INIS)

    Schroeder, J.A.; Pafford, D.J.; Kelly, D.L.; Jones, K.R.; Dallman, F.J.

    1991-01-01

    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

  6. Stainless steels in boiling water reactors. Corrosion problems and possible solutions

    International Nuclear Information System (INIS)

    Combrade, P.; Desestret, A.; Leroy, F.; Coriou, H.

    1977-01-01

    In boiling water reactors, the heat-carrying water may have an up to 0.1 or even 0.2 ppm oxygen content, which can make it highly agressive at operating temperature for stainless steels subject to high physical stresses. Several metallurgical solutions can be considered, and in particular the use of stainless steels having a mixed austenitic-ferritic structure or of standard austenitic steels (18.10 or 18.10 Mo, such as AISI 304 and 316) with carefully controlled carbon and alloy element contents. The behavior of these steels during prolonged tests in water at 288 0 C with a 30 and even 100 ppm oxygen content turned out to be quite satisfactory [fr

  7. Cavitational boiling of liquids

    International Nuclear Information System (INIS)

    Kostyuk, V.V.; Berlin, I.I.; Borisov, N.N.; Karpyshev, A.V.

    1986-01-01

    Transition boiling is a term usually denoting the segment of boiling curve 1-2, where the heat flux, q, decreases as the temperature head, ΔT/sub w/=T/sub w/-T/sub s/, increases. Transition boiling is the subject of numerous papers. Whereas most researchers have studied transition boiling of saturated liquids the authors studied for many years transition boiling of liquids subcooled to the saturation temperature. At high values of subcooling, ΔT/sub sub/=T/sub s/-T/sub 1/, an anomalous dependence of the heat flux density on the temperature head was detected. Unlike a conventional boiling curve, where a single heat flux maximum occurs, another maximum is seen in the transition boiling segment, the boiling being accompanied by strong noise. The authors refer to this kind of boiling as cavitational. This process is largely similar to noisy boiling of helium-II. This article reports experimental findings for cavitational boiling of water, ethanol, freon-113 and noisy boiling of helium-II

  8. Comparative performance of five Mexican plancha-type cookstoves using water boiling tests

    Directory of Open Access Journals (Sweden)

    Paulo Medina

    Full Text Available While plancha-type cookstoves are very popular and widely disseminated in Latin America, few peer review articles exist documenting their detailed technical performance. In this paper we use the standard Water Boiling Tests (WBT to assess the energy and emission performance of five plancha-type cookstoves disseminated in about 450 thousand Mexican rural homes compared to the traditional 3-stone fire (TSF. In the high-power phase, average modified combustion efficiencies (MCE for plancha-type stoves were 97±1% which was higher than TSF 93±4%, and reductions in CO and PM2.5 total emissions were on average 44%. Time to boil and specific fuel consumption, however, were increased in plancha-type stoves compared to the open fire as a result of the reduced overall thermal efficiency of the plancha during WBT. In the simmering phase, plancha-type stoves showed much more consistent performance reductions compared to the TSF. MCE for plancha stoves were on average 98±1% and 95±3% for the TSF, while reductions in CO and PM2.5 total emissions were on average 55%. In this phase 27% average savings in fuel use are achieved by plancha-type stoves. Removal of the plancha rings resulted in savings of specific fuel consumption (SFC, thermal efficiency (TE, and time to boil; however, CO and PM2.5 emissions increased significantly as flue air is drawn through the comal surface rather than through the combustion zone, resulting in suboptimal combustion conditions.International Workshop Agreement (IWA energy performance Tiers for plancha-type stoves ranged from 0 to 1. However, these results contrast sharply with the well documented reductions in fuel consumption during daily cooking activities achieved by these stoves. IWA indoor emissions Tiers are 4 for both PM2.5 and CO using locally measured values for fugitive emissions. Optimization of combustion chamber design on these stoves in Mexico is desirable to further reduce indoor emissions and to reduce the

  9. Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

    International Nuclear Information System (INIS)

    Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

    1992-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-01

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

  11. Applied methods for mitigation of damage by stress corrosion in BWR type reactors

    International Nuclear Information System (INIS)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C.

    1998-01-01

    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)

  12. Characterization of the TIP4P-Ew water model: vapor pressure and boiling point.

    Science.gov (United States)

    Horn, Hans W; Swope, William C; Pitera, Jed W

    2005-11-15

    The liquid-vapor-phase equilibrium properties of the previously developed TIP4P-Ew water model have been studied using thermodynamic integration free-energy simulation techniques in the temperature range of 274-400 K. We stress that free-energy results from simulations need to be corrected in order to be compared to the experiment. This is due to the fact that the thermodynamic end states accessible through simulations correspond to fictitious substances (classical rigid liquids and classical rigid ideal gases) while experiments operate on real substances (liquids and real gases, with quantum effects). After applying analytical corrections the vapor pressure curve obtained from simulated free-energy changes is in excellent agreement with the experimental vapor pressure curve. The boiling point of TIP4P-Ew water under ambient pressure is found to be at 370.3+/-1.9 K, about 7 K higher than the boiling point of TIP4P water (363.7+/-5.1 K; from simulations that employ finite range treatment of electrostatic and Lennard-Jones interactions). This is in contrast to the approximately +15 K by which the temperature of the density maximum and the melting temperature of TIP4P-Ew are shifted relative to TIP4P, indicating that the temperature range over which the liquid phase of TIP4P-Ew is stable is narrower than that of TIP4P and resembles more that of real water. The quality of the vapor pressure results highlights the success of TIP4P-Ew in describing the energetic and entropic aspects of intermolecular interactions in liquid water.

  13. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

    2014-01-01

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Sulatsky, A.A.; Khabensky, V.B. [Sci. Res. Technol. Inst., Leningrad (Russian Federation); Lopukh, D.B.; Petrov, Y.B.; Pechenkov, A.Y. [St. Petersburg Electrotechnical University (SPbEU), Prof. Popov st 5/3, St. Petersburg (Russian Federation)

    2000-01-01

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the 'Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO{sub 2+x}-16% ZrO{sub 2}-15% Fe{sub 2}O{sub 3}-6% Cr{sub 2}O{sub 3}-3% Ni{sub 2}O{sub 3}. The melt surface temperature ranged within 1920-1970 K. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V. [Research Institute of Technology, Sosnovy Bor (NITI) (RU)] [and others

    1999-07-01

    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%UO{sub 2}- 16%ZrO{sub 2}- 15%Fe{sub 2}O{sub 3} - 6%Cr{sub 2}O{sub 3}-3%Ni{sub 2}O{sub 3}. The melt surface temperature was 1650-1700degC. (author)

  16. Possibilities of crack mouth opening displacement (CMOD) measurement under boiling and pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Ehling, W.

    1984-01-01

    Fracture mechanics investigations carried out so far in laboratory conditions cover only part of the material stresses, as effects which occur in nuclear powerstations, in particular, such as corrosion and radioactive radiation are largely left out of account. Therefore experiments including these effects were recently carried out in autoclaves, test rigs simulating reactors (HRD experimental plant) and in experimental reactors. An important parameter of experimental fracture mechanics is the measurement of crack opening displacement (COD). The crack opening is measured with socalled clip gauges (transmitters based on strain gauges, which convert mechanical deformation of springs into electrical signals) on standard samples in the laboratory. It was therefore sensible to use these high temperature strain gauges (HTD) for the development of a measuring system for travel for pressurized water and boiling water reactor conditions. (orig.) [de

  17. Limiting factor analysis of high availability nuclear plants (boiling water reactors). Final report

    International Nuclear Information System (INIS)

    Frederick, L.G.; Brady, R.M.; Shor, S.W.W.; McCusker, J.T.; Alden, W.M.; Kovacs, S.

    1979-08-01

    The pertinent results are presented of a 16-month study conducted for Electric Power Research Institute by General Electric Company, Bechtel Power Corporation, and Philadelphia Electric Company. The study centered around the Peach Bottom 2 Atomic Power Station, but also included limited study of operations at 20 additional operating boiling water reactors. The purpose of the study was to identify and evaluate key factors limiting plant availability, and to identify potential improvements for eliminating or alleviating those limitations. The key limiting factors were found to be refueling activities; activities related to the reactor fuel; reactor scrams; activities related to 20 operating systems or major components; delays due to radiation, turbid water during refueling operations, facilities/working conditions, and dirt/foreign material; and general maintenance/repair of valves and piping. Existing programs to reduce the effect on plant unavailability are identified, and suggestions for further action are made

  18. 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-To