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Sample records for bwr instabilities applications

  1. A New Method for Early Anomaly Detection of BWR Instabilities

    International Nuclear Information System (INIS)

    Ivanov, K.N.

    2005-01-01

    The objective of the performed research is to develop an early anomaly detection methodology so as to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The research utilizes a model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, which is used as a generator of time series data for anomaly detection at an early stage. The model captures critical nonlinear features of coupled thermal-hydraulic and nuclear reactor dynamics and (slow time-scale) evolution of the anomalies as non-stationary parameters. The time series data derived from this nonlinear non-stationary model serves as the source of information for generating the symbolic dynamics for characterization of model parameter changes that quantitatively represent small anomalies. The major focus of the presented research activity was on developing and qualifying algorithms of pattern recognition for power instability based on anomaly detection from time series data, which later can be used to formulate real-time decision and control algorithms for suppression of power oscillations for a variety of anticipated operating conditions. The research being performed in the framework of this project is essential to make significant improvement in the capability of thermal instability analyses for enhancing safety, availability, and operational flexibility of currently operating and next generation BWRs.

  2. A New Methodology for Early Anomaly Detection of BWR Instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, K. N.

    2005-11-27

    The objective of the performed research is to develop an early anomaly detection methodology so as to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The research utilizes a model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, which is used as a generator of time series data for anomaly detection at an early stage. The model captures critical nonlinear features of coupled thermal-hydraulic and nuclear reactor dynamics and (slow time-scale) evolution of the anomalies as non-stationary parameters. The time series data derived from this nonlinear non-stationary model serves as the source of information for generating the symbolic dynamics for characterization of model parameter changes that quantitatively represent small anomalies. The major focus of the presented research activity was on developing and qualifying algorithms of pattern recognition for power instability based on anomaly detection from time series data, which later can be used to formulate real-time decision and control algorithms for suppression of power oscillations for a variety of anticipated operating conditions. The research being performed in the framework of this project is essential to make significant improvement in the capability of thermal instability analyses for enhancing safety, availability, and operational flexibility of currently operating and next generation BWRs.

  3. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  4. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    Alessandro Petruzzi; Shin Chin; Kostadin Ivanov; Asok Ray; Fan-Bill Cheung

    2005-01-01

    Full text of publication follows: The objective of the work presented in this paper research, which is supported by the US Department of Energy under the NEER program, is to develop an early anomaly detection methodology in order to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, based on the US NRC coupled code TRACE/PARCS, is being utilized as a generator of time series data for anomaly detection at an early stage. The concept of the methodology is based on the fact that nonlinear systems show bifurcation, which is a change in the qualitative behavior as the system parameters vary. Some of these parameters may change on their own accord and account for the anomaly, while certain parameters can be altered in a controlled fashion. The non-linear, non-autonomous BWR system model considered in this research exhibits phenomena at two time scales. Anomalies occur at the slow time scale while the observation of the dynamical behavior, based on which inferences are made, takes place at the fast time scale. It is assumed that: (i) the system behavior is stationary at the fast time scale; and (ii) any observable non-stationary behavior is associated with parametric changes evolving at the slow time scale. The goal is to make inferences about evolving anomalies based on the asymptotic behavior derived from the computer simulation. However, only sufficient changes in the slowly varying parameter may lead to detectable difference in the asymptotic behavior. The need to detect such small changes in parameters and hence early detection of an anomaly motivate the utilized stimulus-response approach. In this approach, the model

  5. Effect of the inlet throttling on the thermal-hydraulic instability of the natural circulation BWR

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Inada, Fumio; Yoneda, Kimitoshi

    1997-01-01

    Although it is well-established that inlet restriction has a stabilizing for forced circulation BWR, the effect of inlet on the thermal-hydraulic stability of natural circulation BWR remains unknown since increasing inlet restriction affect thermal-hydraulic stability due to reduction of the recirculation flow rate. Therefore experiments have been conducted to investigate the effect of inlet restriction on the thermal-hydraulic stability. A test facility used in this experiments was designed and constructed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation was described as a function of heat flux and inlet subcooling independent of inlet restriction. Stability maps in reference to the channel inlet subcooling, heat flux were presented for various inlet restriction which were carried out by an analysis based on the homogeneous flow various using this function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux. (author)

  6. Experimental study on thermo-hydraulic instability on reduced-moderation natural circulation BWR concept

    International Nuclear Information System (INIS)

    Watanabe, Noriyuki; Subki, M.H.; Kikura, Hiroshige; Aritomi, Masanori

    2003-01-01

    Reduced-moderation natural circulation BWR has been promoted to solve the recent challenges in BWR nuclear power technology problems as one of advanced small and medium-sized reactors equipped with the passive safety features in conformity with the natural law. However, the elimination of recirculation pumps and a high-density core due to the increase of conversion ratio could cause various thermo-hydraulic instabilities especially during the start-up stage. The occurrences of the thermo-hydraulic instabilities are not desirable and it is one of the main challenges in establishing reduced-moderation natural circulation BWR as a commercial reactor. The purpose of this present study is to experimentally investigate the driving mechanism of the thermo-hydraulic instabilities and the effect of system pressure on the unstable flow patterns. Hence, as the fundamental research for this study, a natural circulation loop that carries boiling fluid with parallel boiling channel has been constructed. Channel gap that has been set at 2 mm in order to simulate reduced-moderation reactor core. Pressure ranges of 0.1 up to 0.7 MPa, input heat flux range of 0 ou to 577 kW/m 2 , and inlet subcooling temperatures of 5, 10, and 15 K respectively, are imposed in the experiments. This experiment clarifies that changes in unstable flow patterns with increase in heat flux can be classified into two in response to system pressure range. In case of atmospheric pressure, unstable flow patters has been classified in beyond order, (1) in-phase geysering, (2) transition oscillation combined with both features of in-phase geysering and natural circulation oscillation, (3) natural circulation oscillation induced by hydrostatic head fluctuation, (4) density wave oscillation, and finally (5) stable boiling two-phase flow. On the other hand, in the system pressure range from 0.2 to 0.7 MPa, unstable patters have been dramatically changed in the following order (1) out-of-phase geysering, (2

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. Cause-Effect relationship of the Laguna Verde BWR power instability by empirical mode decomposition

    International Nuclear Information System (INIS)

    Blazquez, J.; Ruiz, J.; Castillo, R.

    2008-01-01

    The signals coming from natural phenomena are in essence non lineal and not stationary. A recent development, well-known as Empirical Mode Decomposition (EMD) it presents a novel focus that allows to represent in adaptative form non stationary signals as a sum of components of half zero. These components denominated Intrinsic Mode Functions (IMF) they help to the analysis of the frequency composition of unidimensional signals. The use of the EMD followed by the Hilbert transform of the IMFs it allows to carry out an analysis in time-frequency of the non lineal and not stationary data. This technique is known as the Hilbert Huang Transform (HHT). In this work a power instability event occurred in January 24, 1995 in the unit I of the nuclear power station of Laguna Verde (Mexico), corresponding to a BWR/5 is analyzed. When a Nuclear Plant suffers a power instability event, it is required obligatorily to explain to the Regulator Organism the effects and the causes of the event. The effects are described simply; not in vain there is a registration of signals in the Process Computer of where the required information is extracted. But the causes are not always immediate and easy for to identify. The power instability can happen during the start, when the refrigeration flow is relatively low in front to the power. By reason of that the reactivity coefficient by holes is negative, the power oscillates with a very defined frequency, generally of the order of 0.5 Hz. If the oscillations increase progressively of amplitude, we are in an instability event. It is interesting to include in the report the instant in that the began instability and the actions of the operator before and after the same one. As the actions are registered, the investigation is focused toward the instant of the beginning to be able to identify them. In this work the power signal in five empiric ways of Hilbert-Huang and a residual breaks down. The instability is only reflected in the way of smaller

  9. The Physical Mechanism of Core-Wide and Local Instabilities at the Forsmark-1 BWR

    International Nuclear Information System (INIS)

    Analytis, G. Th.

    1998-10-01

    During the last 15 years, the problem of BWR instabilities has attracted the attention of a number of researchers. From the theoretical point of view, one would be interested in physically understanding the mechanisms responsible for the in- and out-of-phase core wide power oscillations observed at certain operating points of the power-flow map in different BWRs. From the practical point of view, one must try to avoid these 'incidents' since either locally, or globally, the power may substantially exceed the prescribed levels. In this work, we shall use RAMONA3-12 and analyse a rather unusual instability incident at Forsmark-1 in which in addition to the core-wide fundamental spatial mode oscillation, there were local large amplitude power oscillations at different radial positions in the core. We were able to reproduce these unusual experimental findings by assuming that there are large amplitude Density Wave Oscillations (DWOs) in different bundles, induced by the fact that these bundles were not seated properly into the lower fuel support plate. (author)

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

    Directory of Open Access Journals (Sweden)

    Antonella Lombardi Costa

    2008-01-01

    Full Text Available Boiling water reactor (BWR instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presented. The RELAP5/MOD3.3 thermal-hydraulic (TH system code and the PARCS-2.4 3D neutron kinetic (NK code were coupled to simulate BWR transients. Different algorithms were used to calculate the decay ratio (DR and the natural frequency (NF from the power oscillation predicted by the transient calculations as two typical parameters used to provide a quantitative description of instabilities. The validation of the code model set up for the Peach Bottom Unit 2 BWR plant is performed against low-flow stability tests (LFSTs. The four series of LFST have been performed during the first quarter of 1977 at the end of cycle 2 in Pennsylvania. The tests were intended to measure the reactor core stability margins at the limiting conditions used in design and safety analyses.

  11. Using the Hurst's exponent as a monitor and predictor of BWR reactor instabilities

    International Nuclear Information System (INIS)

    Gavilan Moreno, Carlos J.

    2010-01-01

    Since the decade of the 1950s, when the development of boiling water reactor technology began, unstable situations have existed, which involve a high amplitude self-oscillatory process in the reactor's thermal power. As the development progressed and the reactors increased power density, the possibility of instability under certain circumstances increased. Thus, in 1985, Caorso nuclear plant (Italy) reported the first event of this type, and in 1988, such an event was reported at La Salle as well. Since then, multiple instability events have been reported. The danger of these unstable power situations resides in the possibility of exceeding a thermal limit, as expressed in Appendix A of 10FR50. Thus, the need arises to monitor and correct these situations in the industry. The most common way to monitor and control these instability situations involves the use of Decay Ratio (DR) and Resonance Frequency (RF). The use of these parameters is polemical, because their use involves certain simplifications and operations prior to the calculation which question how well they represent the reality. The most important simplifications are those which lead to the interpretation of the power time series as the result of a second order system. With regard to the previous operations, the time series needs to be standardized and filtered. The result is loss of information during prediction, due to the operations, and the results, therefore, lack accuracy. In this paper, the system is considered without simplifications, that is to say that it is treated as dynamic and, as we shall see, chaotic, in the mathematical sense of the term. The series will be used in pure form without manipulations. The parameter used for monitoring and prediction of the core's behaviour will be the Hurst's exponent (H). The concept used for this proposal is that the response of a complex dynamic system depends not only on the last excitation, but on the prior history. The processes and systems are

  12. Application of TRAB to BWR transient and accident analyses

    International Nuclear Information System (INIS)

    Raety, H.

    1995-01-01

    TRAB is a one-dimensional reactor dynamics code developed at VTT (Technical Research Centre of Finland) for BWRs. It models the core and the main circulation system inside the reactor vessel, including the steam dome with related systems, steam lines, recirculation pumps, incoming and outgoing flows as well as control and protection systems. The core model includes a one-dimensional description of the geometry, neutronics, rod heat transfer, and thermal hydraulics, using presently at most three parallel axial hydraulic channels. A synthesis model in the neutronics allows to include radially asymmetric phenomena in the modelling. TRAB has been extensively used for analyses of the TVO ABB Atom type BWRs in Olkiluoto, including analyses of pump trip, overpressurization, rod drop, ATWS, and stability. Analyses by VTT have been carried out both for the TVO Power Company and the Finnish Centre for Radiation and Nuclear Safety (STUK). The code is also used for BWR analyses at STUK. (orig.) (12 refs., 5 figs., 1 tab.)

  13. Application of eddy current inspection to the Inconel weld of BWR internals

    International Nuclear Information System (INIS)

    Machida, Eiji; Yusa, Noritaka

    2004-01-01

    In order to definite the basic specifications of application of ECT (Eddy Current Test) to Inconel weld of BWR internals, the inspection and numerical analysis were carried out. The characteristics of the existing ECT probe were studied by making sample as same as CRD stud tube, measuring the relative permeability and electric conductivity of Inconel and alloy and evaluating ECT probe. On the basis of the results obtained, the basic specifications were determined and a new eddy current probe for inspection was designed and produced. The new ECT probe was able to detect small notch in Inconel weld, to classify the defects by eddy current inspection signal and sizing the length and depth. It is concluded that the new ECT probe is able to apply the Inconel weld of BWR internals. (S.Y.)

  14. Application of process computers and colour CRT displays in the plant control room of a BWR

    International Nuclear Information System (INIS)

    Itoh, M.; Hayakawa, H.; Kawahara, H.; Neda, T.; Wakabayashi, Y.

    1983-01-01

    The recent application of a CRT display system in an 1100-MW(e) BWR plant control room and the design features of a new control room whose installation is planned for the next generation are discussed. As reactor unit capacity and the need for plant safety and reliability continue to increase, instrumentation and control equipment is growing in number and complexity. In consequence, control and supervision of plant operations require improvement. Thus, because of recent progress in the field of process computers and display equipment (colour CRTs), efficient improvements of the control room are under way in the Japanese BWR plant. In the recently constructed BWR plant (1100 MW(e)), five CRTs on the bench board and two process computers were additionally installed in the control room during the construction stage to improve plant control and supervisory functions by implementing the lessons learned from the Three Mile Island incident. The major functions of the new computers and display systems are to show integrated graphic displays of the plant status, to monitor the standby condition of the safety system, to show the condition of the integrated alarm system, etc. In practice, in the actual plant, this newly installed system performs well. On the basis of the experience gained in these activities, a new computerized control and monitoring system is now being designed for subsequent domestic BWR plants. This advanced system will incorporate not only the functions already mentioned, but also a surveillance guide system and plant automation. For future plants, a diagnostic system and an instructional system that can analyse a disturbance and give operational guidance to the plant operator are being developed in a government-sponsored programme. (author)

  15. Cause-Effect relationship of the Laguna Verde BWR power instability by empirical mode decomposition; Relacion efecto-causa de la inestabilidad de potencia del BWR de Laguna Verde por descomposicion modal empirica

    Energy Technology Data Exchange (ETDEWEB)

    Blazquez, J.; Ruiz, J.; Castillo, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2008-07-01

    The signals coming from natural phenomena are in essence non lineal and not stationary. A recent development, well-known as Empirical Mode Decomposition (EMD) it presents a novel focus that allows to represent in adaptative form non stationary signals as a sum of components of half zero. These components denominated Intrinsic Mode Functions (IMF) they help to the analysis of the frequency composition of unidimensional signals. The use of the EMD followed by the Hilbert transform of the IMFs it allows to carry out an analysis in time-frequency of the non lineal and not stationary data. This technique is known as the Hilbert Huang Transform (HHT). In this work a power instability event occurred in January 24, 1995 in the unit I of the nuclear power station of Laguna Verde (Mexico), corresponding to a BWR/5 is analyzed. When a Nuclear Plant suffers a power instability event, it is required obligatorily to explain to the Regulator Organism the effects and the causes of the event. The effects are described simply; not in vain there is a registration of signals in the Process Computer of where the required information is extracted. But the causes are not always immediate and easy for to identify. The power instability can happen during the start, when the refrigeration flow is relatively low in front to the power. By reason of that the reactivity coefficient by holes is negative, the power oscillates with a very defined frequency, generally of the order of 0.5 Hz. If the oscillations increase progressively of amplitude, we are in an instability event. It is interesting to include in the report the instant in that the began instability and the actions of the operator before and after the same one. As the actions are registered, the investigation is focused toward the instant of the beginning to be able to identify them. In this work the power signal in five empiric ways of Hilbert-Huang and a residual breaks down. The instability is only reflected in the way of smaller

  16. Theory and control calculations. Application to BWR and PWR

    International Nuclear Information System (INIS)

    Velarde, G.; Guerra, J.

    1972-01-01

    A theoretical method to extend the applicability of P 1 approximation to strong absorbers is given. Tables of effective diffusion parameters for slab and cylindrical geometry as well as a summary of expressions for obtaining the extrapolation length in cases of interest, are included. (Author) 21 refs

  17. BWR Stability Issues in Japan

    Directory of Open Access Journals (Sweden)

    Hideaki Ikeda

    2008-01-01

    Full Text Available The present paper reviews activates relevant to the boiling water reactor (BWR stability phenomenon, which has a coupled neutronic and thermal-hydraulic nature, from the viewpoint of model and code developments and their applications to the BWR stability solution methodology in Japan.

  18. An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Masashi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-07-01

    This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

  19. Application of the IBERDROLA RETRAN Licensing Methodology to the Confrentes BWR-6 110% Extended Power Uprate

    International Nuclear Information System (INIS)

    Fuente, Rafael de la; Iglesias, Javier; Sedano, Pablo G.; Mata, Pedro

    2003-01-01

    IBERDROLA (Spanish utility) and IBERDROLA INGENIERIA (engineering branch) have been developing during the last 2 yr the 110% Extended Power Uprate Project for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved in advance by the Spanish Nuclear Regulatory Authority. This methodology has been applied to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 and 13 and to develop a significant number of safety analyses of the Cofrentes Extended Power.Because the scope of the licensing process of the Cofrentes Extended Power Uprate exceeds the range of analysis included in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes RETRAN model to the analysis of new transients. This is the case of the total loss of feedwater (TLFW) transient.The content of this paper shows the benefits of having an in-house design and licensing methodology and describes the process to extend the applicability of the Cofrentes RETRAN model to the analysis of new transients, particularly in this paper the TLFW transient

  20. Application of version 3.1 of EPRI BWR radiolysis model

    International Nuclear Information System (INIS)

    Version 3.1 of the EPRI BWR vessel internals application (BWRVIA) code for calculating oxidant and electrochemical corrosion potential (ECP) around a BWR primary circuit has recently been released and this paper outlines the changes that have been carried out to the model and how the model compares with plant observations. There were two primary motivations for the development of BWRVIA V3.1 for plants injecting hydrogen into the feedwater to mitigate intergranular stress corrosion cracking (IGSCC) of reactor piping and internals; the fact that many BWRs now add Pt to the primary system to catalyze hydrogen:oxidant recombination at surfaces so the model needs to provide an accurate description of molar ratio (ratio of hydrogen to oxidant) around the primary circuit, and secondly to improve predictions of ECP in the lower plenum region for plants operating under moderate hydrogen water chemistry (HWC-M). Version 3.1 upgraded the model's benchmark for neutron and gamma dose rates and provided for model calculations with core axial power shapes that were bottom, middle and top peaked, characteristic of some core designs at beginning, middle, and end of cycle conditions. Improved reaction rate expressions also were incorporated along with refinements based on sensitivity testing and comparison to plant data under noble metal hydrogen water chemistry regimes. In the presence of Pt deposits on surfaces, molar ratios greater than 2 at a particular location in the primary circuit imply reducing conditions, low ECP and therefore protection from stress corrosion cracking. Plants that apply noble metal will therefore be protected from SCC in these locations. In recent years several HWC-M plants have obtained ECP data from local power range monitors sampling water from the bottom head of the vessel. These ECP measurements have shown that not all BWRs respond similarly to hydrogen addition with some plants requiring very high feed water hydrogen levels to achieve ECP

  1. BWR containment vessel drywell head bolt de-tensioning during noble metal chemical application

    International Nuclear Information System (INIS)

    Herrera, M.; Mattson, R.

    2001-01-01

    Implementation of Noble Metal Chemical Application (NMCA) in a Boiling Water Reactor (BWR) requires injection of a noble metal compound while the reactor is idle at hot shutdown conditions. In order to minimize outage time, utilities are very pro-actively finding ways to reduce the number of critical path tasks. One of these tasks is to remove some containment vessel drywell head bolts during the NMCA idle time. This, thereby, saves the utilities outage time, or they can perform other tasks as desired. Using design basis conditions and state-of-the-art analytical techniques, detailed finite element stress analyses of the closure region are performed. Two acceptance criteria are evaluated. The first is that contained within Section III of the ASME Code for allowable stresses. The second relates to leak tightness, i.e., with bolt removal the joint must still remain leak tight. This paper describes the activities performed to justify bolt removal during the noble metal compound injection idle time at a particular plant. Drywell joint integrity criteria, design input, and assumptions are defined, and analytical results justifying bolt removal are presented. (author)

  2. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

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

  3. On the fast estimation of transit times application to BWR simulated data

    International Nuclear Information System (INIS)

    Antonopoulos-Domis, M.; Marseguerra, M.; Padovani, E.

    1996-01-01

    Real time estimators of transit times are proposed. BWR noise is simulated including a global component due to rod vibration. The time obtained form the simulation is used to investigate the robustness and noise immunity of the estimators. It is found that, in presence of a coincident (global) signal, the cross-correlation function is the worst estimator. (authors)

  4. Design-by-analysis application to next generation BWR fuel thermal design method

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Mitsutake, Toru

    2008-01-01

    For BWR core thermal design methods, the full mock-up thermal-hydraulic tests have been needed, such as the critical power measurement test, pressure drop measurement test and so on for more than thirty years since 1976. However, the full mock-up test required the high costs and large-scale test facility. Now, there are only a few test facilities in the world to perform the full mock-up thermal-hydraulic test. Moreover, for future BWR, the bundle size tends to be larger, because of reducing the plant construction costs and minimizing days taken for the periodic check. For instance, AB1600, improved ABWR, was proposed from Toshiba (Imaruoka et al., 2005), whose bundle size was 1.2 times larger than the conventional BWR fuel size. It is too expensive and far from realistic to perform the full mock-up thermal-hydraulic test for such a large size fuel bundle on every new bundle design. The new design procedure is exactly expected at reasonable costs and required to perform the large scale bundle design development, especially for the next generation BWR. Then, we propose a kind of Design by Analysis method combined with the alternative to the full mock-up tests. We called Practical Design-by-Analysis (PDBA) method even more effective and with better accuracy than Full Design-by-Analysis (FDBA). The PDBA method was applied successfully to produce the critical power correlation for the current 8x8 fuel. The estimation error was almost equal to that of the design correlation developed empirically with the full mock-up test data. (author)

  5. Technical Specification action statements requiring shutdown. A risk perspective with application to the RHR/SSW systems of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy, Espoo (Finland); Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States)

    1993-11-01

    When safety systems fail during power operation, the limiting conditions for operation (LCOs) and associated action statements of technical specifications typically require that the plant be shut down within the limits of allowed outage time (AOT). However, when a system needed to remove decay heat, such as the residual heat removal (RHR) system, is inoperable or degraded, shutting down the plant may not necessarily be preferable, from a risk perspective, to continuing power operation over a usual repair time, giving priority to the repairs. The risk impact of the basic operational alternatives, i.e., continued operation or shutdown, was evaluated for failures in the RHR and standby service water (SSW) systems of a boiling-water reactor (BWR) nuclear power plant. A complete or partial failure of the SSW system fails or degrades not only the RHR system but other front-line safety systems supported by the SSW system. This report presents the methodology to evaluate the risk impact of LCOs and associated AOT; the results of risk evaluation from its application to the RHR and SSW systems of a BWR; the findings from the risk-sensitivity analyses to identify alternative operational policies; and the major insights and recommendations to improve the technical specifications action statements.

  6. Practical application of current fluid/structure interaction technology to the design and analysis to BWR containment vessels

    International Nuclear Information System (INIS)

    Edwards, N.W.

    1979-01-01

    The containment systems for all BWR reactors consist of a drywell which contains the reactor and a wetwell which contains a suppression pool. The wetwell structure and the suppression pool are clearly condidates for the application of current - if not state-of-the-art - fluid/structure interaction technology. The original design of the early BWR containments gave little consideration to the pool hydrodynamic loads themselves, to say nothing of the fluid/structure interaction aspects of the phenomena. In later years, loads were specified to act on the walls of the suppression chamber, but it was left to the designer as to how he would account for the mass of water within the suppression pool. The state-of-the-art is rapidly approaching the point where it will be possible to consider the pool coupled to the structure with a pressure source acting in the pool. However, one must exercise caution in order to make practical use of this developing technology. (orig.)

  7. Alternative Zr alloys with irradiation resistant precipitates for high burnup BWR application

    International Nuclear Information System (INIS)

    Garzarolli, F.; Ruhmann, H.; Van Swan, L.

    2002-01-01

    In the core of BWRs, the second-phase particles (SPP) of Zircaloy-2 and Zircaloy-4, the Zr(FeCr) 2 and the Zr 2 (FeNi) phase, release Fe and dissolve. The degree of dissolution depends on initial size and fluence. These SPP, however, are important for the corrosion behavior of Zircaloy. Zircaloy shows an increase of corrosion at a certain burnup, depending on the initial SPP size and fast neutron fluence. Only Zr alloys with irradiation resistant SPP avoid this type of increased corrosion completely. Two types of irradiation resistant materials were considered. One is a Zr-Sn-Fe alloy containing the Zr 3 Fe phase, which is irradiation resistant under BWR conditions. The other material is a Zr-Sn-Nb alloy containing the irradiation resistant β-Nb phase. In-BWR tests have shown that a Sn content of >0.8% is mandatory to minimize the nodular corrosion. Two prototypes of irradiation resistant alloys, Zr1.3Sn0.25-0.3 Fe and Zr1Sn2-3Nb, were irradiated in a BWR for 1372 days to a fast fluence of 9 x 10 21 n/cm 2 (E > 1 MeV). These irradiation tests showed that Zr1.3Sn0.25-0.3 Fe has a little lower resistance against nodular corrosion than optimized LTP (Low Temperature Process) Zircaloy-2/4 and revealed that Zr1Sn2-3Nb is superior to LTP Zircaloy-2/4 with respect to nodular and shadow corrosion resistance. The BWR corrosion resistance of Zr1Sn2-3Nb depends on heat treatment. The lowest corrosion was observed with material fabricated completely in the α-range, but also material manufactured in the lower (α+β)-range exhibits low corrosion. Material fabricated in the upper (α+β)-range showed a somewhat higher corrosion, a corrosion behavior similar to LTP Zircaloy-2/4. As far as final annealing is concerned, a long time annealing at 540 deg C is superior to a standard recrystallization treatment (e.g., at 580 deg C), which still leads to a corrosion behavior that is better than stress relieved Zr1Sn2-3Nb. Zr1Sn2-3Nb is resistant to shadow corrosion, when fabricated

  8. 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

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-04-01

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

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

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  12. 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

  13. Application Of The Iberdrola Licensing Methodology To The Cofrentes BWR-6 110% Extended Power Up-rate

    International Nuclear Information System (INIS)

    Mata, Pedro; Fuente, Rafael de la; Iglesias, Javier; Sedano, Pablo G.

    2002-01-01

    Iberdrola (spanish utility) and Iberdrola Ingenieria (engineering branch) have been developing during the last two years the 110% Extended Power Up-rate Project (EPU 110%) for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved by the Spanish Nuclear Regulatory Authority. This methodology has been already used to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 to 14. The methodology has been also applied to develop a significant number of safety analysis of the Cofrentes Extended Power Up-rate including: Reactor Heat Balance, Core and Fuel performance, Thermal Hydraulic Stability, ECCS LOCA Evaluation, Transient Analysis, Anticipated Transient Without Scram (ATWS) and Station Blackout (SBO) Since the scope of the licensing process of the Cofrentes Extended Power Up-rate exceeds the range of analysis included in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes licensing methodology to the analysis of new transients. This is the case of the TLFW transient. The content of this paper shows the benefits of having an in-house design and licensing methodology, and describes the process to extend the applicability of the methodology to the analysis of new transients. The case of analysis of Total Loss of Feedwater with the Cofrentes Retran Model is included as an example of this process. (authors)

  14. Development of quick-response area-averaged void fraction meter. Application to BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-05-01

    Authors have been developed a practical conductance-type void fraction meter to measure instantaneously area-averaged void fraction in rod bundle. The principle of the meter is based on the fact that the electrical conductance changes with the change of void fraction in gas-liquid two-phase flow. According to air/water two-phase flow experiment, the void fraction was approximated by {alpha}=1-I/I{sub 0}, where {alpha} and I are void fraction and current (I{sub 0} is current at {alpha}=0). Authors investigated the performance of the void fraction meter under high temperature/high pressure conditions (BWR condition; 290degC, 7MPa). The results indicated that the void fraction was approximated by {alpha}=1-I/I{sub 0} even under high temperature/high pressure condition of stem/water flow. However, it is necessary to take account of temperature dependency of water specific conductance. Therefore, authors derived a correction equation for temperature dependency. Further, for applying the void fraction meter to a large-scale facility, it was found to be necessary to reduce the capacitance of the circuit. Then, authors developed the method to reduce the capacitance effect. Finally, authors succeeded to measure the void fraction in 2 x 2 bundle flow path at the range of 0% - 70% in the error of 10% under high temperature/high pressure and mass flux of less than 133 kg/m{sup 2}s. Developed void fraction meter is theoretically not affected by flow rate. Therefore, it can be applied to the condition of oscillating flow. (author)

  15. 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)

  16. Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

    International Nuclear Information System (INIS)

    Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

    2006-01-01

    The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

  17. Theory and control calculations. Application to BWR and PWR; Teoria y calculo de elementos de control, Aplicacion a PWR y BWR

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, G.; Guerra, J.

    1972-07-01

    A theoretical method to extend the applicability of P{sub 1} approximation to strong absorbers is given. Tables of effective diffusion parameters for slab and cylindrical geometry as well as a summary of expressions for obtaining the extrapolation length in cases of interest, are included. (Author) 21 refs.

  18. Detecting long-range correlation with detrended fluctuation analysis: Application to BWR stability

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, Gilberto [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Apartado Postal 55-534, Mexico, DF 09340 (Mexico)]. E-mail: gepe@xanum.uam.mx; Alvarez-Ramirez, Jose [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Apartado Postal 55-534, Mexico, DF 09340 (Mexico); Vazquez, Alejandro [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Apartado Postal 55-534, Mexico, DF 09340 (Mexico)

    2006-11-15

    The aim of this paper is to explore the application of detrended fluctuation analysis (DFA) to study boiling water reactor stability. DFA is a scaling method commonly used for detecting long-range correlations in non-stationary time series. This method is based on the random walk theory and was applied to neutronic power signal of Forsmark stability benchmark. Our results shows that the scaling properties breakdown during unstable oscillations.

  19. Consistency considerations in the use of point kinetics for BWR application

    International Nuclear Information System (INIS)

    Holzer, J.M.; Habert, R.; Pilat, E.E.

    1981-01-01

    The basic question of producing point reactivity parameters for use in RETRAN anaylses is addressed. The technique used in establishing a methodology consists of a stepwise reduction of resolution, in space and time, so as to identify possible areas in which error may be induced and to establish procedures that retain consistency and accuracy. The presented calculational flow plan culminating from this analysis will ultimately be used at Yankee Atomic Electric for design application

  20. 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

  1. BWR crack control

    International Nuclear Information System (INIS)

    Immel, R.

    1981-01-01

    Cooperative research by the Electric Power Research Institute (EPRI) and the Boiling Water Reactor Owners Group (BWROG) may be finding a solution to the problem of stress corrosion cracking in reactor piping, which had caused 254 cracks in BWRs around the world by 1981. Welded joints of type-304 stainless steel in BWR recirculation systems are particularly vulnerable to microscopic stress corrosion cracking, which differs from fatigue cracking because it occurs when electrochemical processes break down the protective film in the zones affected by welding heat. The cracking is not hazardous to the public because it takes place within the containment building and causes leaks that are quickly detected. About one percent of welds crack, causing downtime and personnel exposure. The EPRI-BWROG project used a model of stress, water condition, and steel sensitization to find out the causes, develop remedies, and get the remedies into the field. Among the remedies are the adaptive learning network, an automated, microprocessor-controlled ultrasonic testing system that is scheduled for field testing and utility application in 1982. 2 figures

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  4. Application of Detailed Chemical Kinetics to Combustion Instability Modeling

    Science.gov (United States)

    2016-01-04

    sustained by combustion . Irreparable damage can occur in less than 1 second. CI caused a four year delay in the development of the F-1 engine used...Distribution A: approved for public release; distribution unlimited. 2 Challenges of Combustion Instability Damaged engine injector faceplate...caused by combustion instability “ Combustion instabilities have been observed in almost every engine development effort, including even the most recent

  5. Development and application of a Fourier transform based methodology for the identification of instability in boiling water reactors at a local scale

    Energy Technology Data Exchange (ETDEWEB)

    Walser, Stefan Franz

    2017-04-06

    This thesis addresses the development of an analysis methodology for BWR instability phenomena and aims at the identification of in-core, local thermal-hydraulic processes during a transient. The analysis methodology is designed to apply as input data the simulation results of time domain coupled system codes. For the application described in this thesis, a coupled TRACE/PARCS model representing the Oskarshamn-2 (O2) NPP with a one-to-one core channel representation has been used. The coupled model simulates the O2-1999 feedwater transient; an instability event characterized by an in-phase mode of oscillation with reactor power amplitudes up to 132 %. The analysis methodology is a two-step approach and uses in the first step the fast-fourier transform algorithm applied on normalized core parameters in twodimensional spacial direction of the core. The normalization of the data implies the advantage of directly comparable results in spectral representation. The spectral analysis results show for each data node the oscillation amplitude to its corresponding frequency. In the second step the dominating frequency of each single parameter is determined and the relative phase shift of the dominating components is calculated. The application of the developed methodology on the simulation results of the O2-1999 feedwater transient show that the channel mass flow rates have among all investigated parameters the clearest differences in the local expression of oscillation and are a governing indicator for BWR instability due to the density wave mechanism. The spectral analysis of the core channel in planar direction points out a heterogeneous oscillation behavior of the fuel assemblies mass fl ow rates. A certain pattern of core channels with striking mass flow rate oscillations is prevailing and the pattern shows proportionality to the fuel assemblies relative power ratio. Moreover the mass ow rate oscillations of the peripheral core channels are observed to employ a quasi

  6. Level-Swell Prediction With RETRAN-3D And Its Application To A BWR Steam-Line-Break Analysis

    International Nuclear Information System (INIS)

    Aounallah, Y.; Hofer, K.

    2003-01-01

    Level-swell experiments have often been simulated using system codes, such as TRAC and RELAP, but only cursory assessments have been performed with the operational-transient code RETRAN-3D, the main system code used within the STARS project. The present study, initiated in the framework of a BWR Steam-Line-Break (SLB) accident scenario, addresses this lacuna by performing RETRAN simulations of the General Electric Level-Swell experiments, and by investigating their implications on power plant accident analyses. Parameters to which the predicted level swell is sensitive have been identified, and recommendations on code options are made. The SLB analysis objective was to determine the amount of steam and liquid discharged through the break under specified boundary conditions, and to gauge the results against reference values. The impact of the nodalization of the upper part of the reactor pressure vessel was investigated and found to play an important role, whereas the level swell induced from flashing was found not to be the predominant factor for these simulations. (author)

  7. BWR stability analysis

    International Nuclear Information System (INIS)

    Valtonen, K.

    1990-01-01

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

  8. Application of autoregressive methods and Lyapunov coefficients for instability studies of nuclear reactors

    International Nuclear Information System (INIS)

    Aruquipa Coloma, Wilmer

    2017-01-01

    Nuclear reactors are susceptible to instability, causing oscillations in reactor power in specific working regions characterized by determined values of power and coolant mass flow. During reactor startup, there is a greater probability that these regions of instability will be present; another reason may be due to transient processes in some reactor parameters. The analysis of the temporal evolution of the power reveals a stable or unstable process after the disturbance in a light water reactor of type BWR (Boiling Water Reactor). In this work, the instability problem was approached in two ways. The first form is based on the ARMA (Autoregressive Moving Average models) model. This model was used to calculate the Decay Ratio (DR) and natural frequency (NF) of the oscillations, parameters that indicate if the one power signal is stable or not. In this sense, the DRARMA code was developed. In the second form, the problems of instability were analyzed using the classical concepts of non-linear systems, such as Lyapunov exponents, phase space and attractors. The Lyapunov exponents quantify the exponential divergence of the trajectories initially close to the phase space and estimate the amount of chaos in a system; the phase space and the attractors describe the dynamic behavior of the system. The main aim of the instability phenomena studies in nuclear reactors is to try to identify points or regions of operation that can lead to power oscillations conditions. The two approaches were applied to two sets of signals. The first set comes from signals of instability events of the commercial Forsmark reactors 1 and 2 and were used to validate the DRARMA code. The second set was obtained from the simulation of transient events of the Peach Bottom reactor; for the simulation, the PARCS and RELAP5 codes were used for the neutronic/thermal hydraulic coupling calculation. For all analyzes made in this work, the Matlab software was used due to its ease of programming and

  9. TRAC-BWR development

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  12. Application of high-speed photography to hydrodynamic instability research

    International Nuclear Information System (INIS)

    Chang Lihua; Li Zuoyou; Xiao Zhengfei; Zou Liyong; Liu Jinhong; Xiong Xueshi

    2012-01-01

    High-speed photography is used to study the Rayleigh-Taylor instability of air-water interface driven by high- pressure exploding gas. Clear images illustrating the instability are obtained, along with the air bubble peak speed and turbulent mixing speed. The RM (Richtmyer-Meshkov) instability of air/SF 6 interface driven by shock wave is also researched by using high-speed Schlieren technique on the horizontal shock tube and primary experimental results are obtained, which show the change of the turbulent mixing region clearly. (authors)

  13. Study of instabilities in phase by using the tool {sup D}ynamics{sup :} analysis of the evolution space temporary of the waves of density in channels of reactors BWR; Estudio de las Inestabilidades en Fase Mediante la Herramienta Dinamics: analisis de la Evolucion Espacio Temporal de las Ondas de Densidad en Canales de Reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Escriva, R.; Merino, R.; Melara, J.

    2013-07-01

    This paper presents the basics of Dynamics V2 to code It allows calculations of stability for oscillations in phase in BWR reactors in the time domain. The equations of the model are exposed and is the integration of the equations. The model can be used in a large number of nodes thrust for the calculations to an acceptable computational cost, it has simplified dynamics of recirculation loop and the code has been incorporated the Oscillation in phase boundary conditions. The code incorporates the equations of boiling sub-cooled which allows to make more realistic calculations as well as subroutines to calculate the subroutines-based properties of the MATPRO and ASME.

  14. 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

  15. 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

  16. The Novel Concept of "Behavioural Instability'' and Its Potential Applications

    DEFF Research Database (Denmark)

    Pertoldi, Cino; Bahrndorff, Simon; Novicic, Zorana Kurbalija

    2016-01-01

    -clockwise (CCW) movement of 19 highly inbred lines of Drosophila melanogaster tested in a circular arena. We show that it is possible to quantify behavioural instability using the indices traditionally used to investigate DI. Results revealed several significant differences among lines, depending on the index...

  17. BWR stability: analysis of cladding temperature for high amplitude oscillations - 146

    International Nuclear Information System (INIS)

    Pohl, P.; Wehle, F.

    2010-01-01

    Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the coolant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / re-wet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes. (authors)

  18. BWR-stability investigation at Forsmark 1

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

    1988-01-01

    A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  19. Growth induced buckling instability of anisotropic tube and its application in wound edge instability

    Directory of Open Access Journals (Sweden)

    Yang Le

    2017-01-01

    Full Text Available Fiber reinforced anisotropic material abounds in biological world. It has been demonstrated in previous theoretical and experimental works that growth of biological soft tubular tissue plays a significant role in morphogenesis and pathology. Here we investigate growth-induced buckling of anisotropic cylindrical tissue, focusing on the effects of type of growth(constraint/unconstraint, isotropic/anisotropic, fiber property(orientation, density and strength, geometry and any interaction between these factors. We studied one-layer and two-layer models and obtained a rich spectrum of results. For one-layer model, we demonstrate that circumferential fiber orientation has a consistent stabilizing effect under various scenarios of growth. Higher fiber density has a destabilizing effect by disabling high-mode buckling. For two-layer model, we found that critical buckling strain at inner boundary is an invariant under same isotropic growth rate ratio between inner/ outer layer(g1 /g0. Then we applied our model to wound healing and illustrate the effects of skin residual stress, fiber property, proliferation region width and wound size on the wound edge stability. We conclude that fiber-reinforcement is an important factor to consider when investigating growth induced instability of anisotropic soft tissue.

  20. Growth induced buckling instability of anisotropic tube and its application in wound edge instability

    Science.gov (United States)

    Yang, Le; Witten, Tarynn M.; Pidaparti, Ramana M.

    2017-01-01

    Fiber reinforced anisotropic material abounds in biological world. It has been demonstrated in previous theoretical and experimental works that growth of biological soft tubular tissue plays a significant role in morphogenesis and pathology. Here we investigate growth-induced buckling of anisotropic cylindrical tissue, focusing on the effects of type of growth(constraint/unconstraint, isotropic/anisotropic), fiber property(orientation, density and strength), geometry and any interaction between these factors. We studied one-layer and two-layer models and obtained a rich spectrum of results. For one-layer model, we demonstrate that circumferential fiber orientation has a consistent stabilizing effect under various scenarios of growth. Higher fiber density has a destabilizing effect by disabling high-mode buckling. For two-layer model, we found that critical buckling strain at inner boundary is an invariant under same isotropic growth rate ratio between inner/ outer layer(g1 /g0). Then we applied our model to wound healing and illustrate the effects of skin residual stress, fiber property, proliferation region width and wound size on the wound edge stability. We conclude that fiber-reinforcement is an important factor to consider when investigating growth induced instability of anisotropic soft tissue.

  1. BWR type nuclear plant

    International Nuclear Information System (INIS)

    Fujita, Yoshio; Okano, Kimifumi; Sasaki, Hiroshi; Okura, Minoru

    1998-01-01

    The present invention provides a BWR plant capable of reducing the size of the reactor building while maintaining reliability of a pool water cooling and cleaning facility even when two fuel storage pools are disposed in the reactor building. Namely, in the reactor building, two fuel storage pools, a temporality storing pool for temporary storing incore structures and a suppression pool are disposed. A primary cleaning facility for cooling and cleaning pool water for each of fuel storage pools comprises a serge tank, a pump, a heat exchanger and a filtration desalting device. A secondary cleaning facility for cleaning pool water in the suppression pool comprises a pump and a filtration desalting device. The first cleaning facility can be switched to be used for the secondary cleaning facility. Specifically, upstream and downstream of the pump of the primary cleaning facility and those of the pump of the secondary cleaning facility are connected by communication pipelines. (I.S.)

  2. 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

  3. 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

  4. 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

  5. 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

  6. Limerick BWR turbine control and protection system upgrade success

    Energy Technology Data Exchange (ETDEWEB)

    Tang, C.K.; Pietryka, T.S.; Federico, P.A., E-mail: tangck@westinghouse.com, E-mail: pietryt@westinghouse, E-mail: federipa@westinghouse.com [Westinghouse Electric Company, LLC, Cranberry Township, PA (United States); Williams, J.C., E-mail: Jonathan.Williams@exeloncorp.com [Exelon Nuclear, Warrenville, IL (United States)

    2015-07-01

    Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

  7. Limerick BWR turbine control and protection system upgrade success

    International Nuclear Information System (INIS)

    Tang, C.K.; Pietryka, T.S.; Federico, P.A.; Williams, J.C.

    2015-01-01

    Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

  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. BWR operation for life extension

    International Nuclear Information System (INIS)

    Stancavage, P.

    1987-01-01

    Most nuclear power plant life extension studies conducted to date focus on the technology, licensing, and economics of extended service life. The most significant factor in plant longevity, however, is proper operation and maintenance. This paper highlights the benefits of boiling water reactor (BWR) operation for life extension and discusses specific recommendations that will enhance the prospects for safe, reliable, and economic power production in the long term. The benefits of BWR operation for life extension include a lower cost of electric energy production, increased capacity and availability factors, a lower forced outage rate, and reduced occupational exposure to radiation. Operating experience and advanced in technology have provided a wealth of knowledge that can be used to develop specific recommendations for adding years to the expected life of BWR plants. This paper discusses key factors in operation for life extension

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

    International Nuclear Information System (INIS)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

  12. Research on combustion instability and application to solid propellant rocket motors. II.

    Science.gov (United States)

    Culick, F. E. C.

    1972-01-01

    Review of the current state of analyses of combustion instability in solid-propellant rocket motors, citing appropriate measurements and observations. The work discussed has become increasingly important, both for the interpretation of laboratory data and for predicting the transient behavior of disturbances in full-scale motors. Two central questions are considered - namely, linear stability and nonlinear behavior. Several classes of problems are discussed as special cases of a general approach to the analysis of combustion instability. Application to motors, and particularly the limitations presently understood, are stressed.

  13. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  14. 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)

  15. Application of Detailed Chemical Kinetics to Combustion Instability Modeling

    Science.gov (United States)

    2016-01-04

    chemical kinetics of the methane oxidation. Two-dimensional results with global chemistry have shown significantly lower amplitudes than the ex...Conference Paper 3. DATES COVERED (From - To) 12 November 2015 – 04 January 2016 4. TITLE AND SUBTITLE Application of Detailed Chemical Kinetics to...under two different conditions corresponding to marginally stable and unstable operation in order to evaluate the performance of the chemical kinetics

  16. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

  17. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  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. Development of next BWR plant

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

  1. Study on the Development of New BWR Core Analysis Scheme Based on the Continuous Energy Monte Carlo Burn-up Calculation Method

    OpenAIRE

    東條, 匡志; tojo, masashi

    2007-01-01

    In this study, a BWR core calculation method is developed. The continuous energy Monte Carlo burn-up calculation code is newly applied to BWR assembly calculations of production level. The applicability of the present new calculation method is verified through the tracking-calculation of commercial BWR.The mechanism and quantitative effects of the error propagations, the spatial discretization and of the temperature distribution in fuel pellet on the Monte Carlo burn-up calculations are clari...

  2. FASMON monitoring of space dependent oscillations in a BWR

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Oguma, R.

    1994-01-01

    BWR-instabilities can occur in fundamental mode with in-phase core wide oscillations or in excited modes as out-of-phase oscillations. The fundamental mode can be successfully monitored with DR (Decay Ratio) and amplitude calculations while the out-of-phase oscillations demands phase analysis between the LPRM-signals in different parts of the core. The present report describes how the BWR-stability monitor FASMON detects such excited modes and presents the topology of the flux swing in a horizontal plane in the core. FASMON is the only available method to detect out-of-phase oscillations fast (within 20 seconds) and robust. The FASMON algorithm is based on adaptive recursive filtering technique. All investigations with FASMON presented in this report show that the software is a good monitor tool. The software makes surveillance of the phase differences and amplitudes for a number of LPRM signals. The method to calculate the phase differences is robust with the following characteristics: The response time within 20 seconds for the FASMON algorithm is very short; Robust - works even under transient conditions; Independent of the signal amplitude; A dead sensor in a LPRM-pair does not cause alarm. The report gives results from investigations using FASMON both on theoretical and measured signals. As a conclusion the investigation of the instability shows that FASMON works even under circumstances with in-phase conditions and low amplitude up to unstable conditions with out-of-phase oscillations and high amplitude. 5 refs, 12 figs

  3. 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).

  4. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning; Perfiles axiales de quemado y fraccion de huecos para calculos de criticidad con credito al quemado para combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-07-01

    This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

  5. Review of two-phase instabilities

    International Nuclear Information System (INIS)

    Kang, Han Ok; Seo, Han Ok; Kang, Hyung Suk; Cho, Bong Hyun; Lee, Doo Jeong

    1997-06-01

    KAERI is carrying out a development of the design for a new type of integral reactors. The once-through helical steam generator is important design features. The study on designs and operating conditions which prevent flow instability should precede the introduction of one-through steam generator. Experiments are currently scheduled to understand two-phase instability, evaluate the effect of each design parameter on the critical point, and determine proper inlet throttling for the prevention of instability. This report covers general two-phase instability with review of existing studies on this topics. The general classification of two phase flow instability and the characteristics of each type of instability are first described. Special attention is paid to BWR core flow instability and once-through steam generator instability. The reactivity feedback and the effect of system parameters are treated mainly for BWR. With relation to once-through steam generators, the characteristics of convective heating and dryout point oscillation are first investigated and then the existing experimental studies are summarized. Finally chapter summarized the proposed correlations for instability boundary conditions. (author). 231 refs., 5 tabs., 47 figs

  6. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning

    International Nuclear Information System (INIS)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-01-01

    This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

  7. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  8. BWR core melt progression phenomena: Experimental analyses

    International Nuclear Information System (INIS)

    Ott, L.J.

    1992-01-01

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

  9. 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)

  10. 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)

  11. Reliable estimation of neutron flux in BWR reactor vessel using the tort code (2) application to neutron and gamma flux estimation

    International Nuclear Information System (INIS)

    Kurosawa, M.; Tsukiyama, T.; Hayashi, K.

    2001-01-01

    A neutron and gamma flux distribution around the core of BWR commercial plant in Japan was calculated, using a three-dimensional transport code, TORT in DOORS32 code system. In the external of the core, the bottom of the model was at an elevation of 150 cm below the bottom of active fuel, the top of the model was at an elevation of the top of the shroud head dome and the radial part of the model was to the outside of the reactor pressure vessel. The top guide beams were modeled explicitly to obtain the neutron and gamma flux distribution both in the beams and outside beams. The each control rod guide tube was also modeled with homogeneous region which included the blade wing and poison tubes so that we could obtain the neutron and gamma flux distribution around the each control rod guide tube. The calculation model mentioned above needed very large memory size which exceeded a few decade giga-bytes. As the using the splicing/coupling method had uncertainly at the splicing/coupling boundary, in this work the calculation was performed without this splicing/coupling method. On the other hand, radioactivity data were measured for a few pieces of the top guide beam, shroud and in-core monitor guide tube in the same plant which was analyzed in the above calculation. So the calculation results were able to be compared with those measured data as benchmarking and at the end of this task, the C/M values at these measured points were obtained and calculation model using TORT was evaluated. (authors)

  12. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  13. 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.

  14. 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

  15. BWR stability analysis with the BNL Engineering Plant Analyzer

    International Nuclear Information System (INIS)

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

    1992-10-01

    March 9, 1989 instability at the LaSalle-2 Power Plant and more than ninety related BWR transients have been simulated on the BNL Engineering Plant Analyzer (EPA). Power peaks were found to be potentially seventeen times greater than the rated power, flow reversal occurs momentarily during large power oscillations, the fuel centerline temperature oscillates between 1,030 and 2,090 K, while the cladding temperature oscillates between 560 and 570 K. The Suppression Pool reaches its specified temperature limit either never or in as little as 4.3 minutes, depending on operator actions and transient scenario. Thermohydraulic oscillations occur at low core coolant flow (both Recirculation Pumps tripped), with sharp axial or redial fission power peaking and with partial loss of feedwater preheating while the feedwater is flow kept high to maintain coolant inventory in the vessel. Effects from BOP system were shown to influence reactor stability strongly through dosed-loop resonance feedback. High feedwater flow and low temperature destabilize the reactor. Low feedwater flow restabilizes the reactor, because of steam condensation and feedwater preheating in the downcomer, which reduces effectively the destabilizing core inlet subcooling. The EPA has been found to be capable of analyzing BWR stability '' shown to be effective for scoping calculations and for supporting accident management

  16. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-22

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

  17. Valuation of power oscillations in a BWR after control rod banks withdrawal events

    International Nuclear Information System (INIS)

    Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F.

    2009-01-01

    The out-of-phase mode of oscillation is a very challenging type of instability occurring in BWR (Boiling Water Reactor) and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, control rod banks (CRB) withdrawal transient was considered to study the power instability occurring in a BWR. To simulate this transient, the control rod banks were continuously removed from the BWR core in different cases. The simulation resulted in a very large increase of power. To perform the instability simulations, the RELAP5/MOD3.3 thermal hydraulic system code was coupled with the PARCS/2.4 3D neutron kinetic code. Data from a real BWR, the Peach Bottom, have been used as reference conditions and reactor parameters. The trend of the mass flow rate, pressure, coolant temperature and the void fraction to four thermal hydraulic channels symmetrically located in the core with respect to the core centre, were taken. It appears that the velocity of the rod bank withdrawal is a very important aspect for reactor stability. The slowest CRB withdrawal (180 s) did not cause power perturbation while the fast removal (20 s) triggered a slow power oscillation that little by little amplified to reach levels of more 100% of the initial power after about 210 s. The investigation of the related thermo hydraulic parameters showed that the mass flow rate, the void fraction and also the coolant temperature began to oscillate at approximately the same time interval

  18. Calculation of BWR [Boiling Water Reactor] limit cycle amplitude using Galerkin's method

    International Nuclear Information System (INIS)

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

    1990-01-01

    This paper describes the application of Galerkin's method to estimate the amplitude of boiling water reactor (BWR) limit cycle oscillations. It will be shown that Galerkin's method can be applied to a model of BWR dynamics consisting of the point kinetics equations and the LAPUR generated feedback transfer function to calculate the time history of small amplitude limit cycles. This allows results from the linear frequency domain code LAPUR to be used to calculate nonlinear time domain information. 2 refs., 2 figs., 1 tab

  19. 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

  20. 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)

  1. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    Tavera D, L.; Camacho L, M.E.

    1991-01-01

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  2. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  3. Natural convection type BWR reactor

    International Nuclear Information System (INIS)

    Tobimatsu, Toshimi.

    1990-01-01

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

  4. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

    Reynolds, R.S.

    1987-01-01

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

  5. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  6. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L.; Camacho L, M.E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  7. 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

  8. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. 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

  10. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  11. NATO Advanced Research Workshop on Chemical Instabilities : Applications in Chemistry, Engineering, Geology, and Materials Science

    CERN Document Server

    Baras, F

    1984-01-01

    On March 14-18, 1983 a workshop on "Chemical Instabilities: Applications in Chemistry, Engineering, Geology, and Materials Science" was held in Austin, Texas, U.S.A. It was organized jointly by the University of Texas at Austin and the Universite Libre de Bruxelles and sponsored qy NATO, NSF, the University of Texas at Austin, the International Solvay Institutes and the Ex­ xon Corporation. The present Volume includes most of the material of the in­ vited lectures delivered in the workshop as well as material from some posters, whose content was directly related to the themes of the invited lectures. In ,recent years, problems related to the stability and the nonlinear dynamics of nonequilibrium systems invaded a great num­ ber of fields ranging from abstract mathematics to biology. One of the most striking aspects of this development is that subjects reputed to be "classical" and "well-established" like chemistry, turned out to give rise to a rich variety of phenomena leading to multiple steady states and...

  12. Cross-section adjustment techniques for BWR adaptive simulation

    Science.gov (United States)

    Jessee, Matthew Anderson

    Computational capability has been developed to adjust multi-group neutron cross-sections to improve the fidelity of boiling water reactor (BWR) modeling and simulation. The method involves propagating multi-group neutron cross-section uncertainties through BWR computational models to evaluate uncertainties in key core attributes such as core k-effective, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multi-group cross-sections to minimize the disagreement between BWR modeling predictions and measured plant data. For this work, measured plant data were virtually simulated in the form of perturbed 3-D nodal power distributions with discrepancies with predictions of the same order of magnitude as expected from plant data. Using the simulated plant data, multi-group cross-section adjustment reduces the error in core k-effective to less than 0.2% and the RMS error in nodal power to 4% (i.e. the noise level of the in-core instrumentation). To ensure that the adapted BWR model predictions are robust, Tikhonov regularization is utilized to control the magnitude of the cross-section adjustment. In contrast to few-group cross-section adjustment, which was the focus of previous research on BWR adaptive simulation, multigroup cross-section adjustment allows for future fuel cycle design optimization to include the determination of optimal fresh fuel assembly designs using the adjusted multi-group cross-sections. The major focus of this work is to efficiently propagate multi-group neutron cross-section uncertainty through BWR lattice physics calculations. Basic neutron cross-section uncertainties are provided in the form of multi-group cross-section covariance matrices. For energy groups in the resolved resonance energy range, the cross-section uncertainties are computed using an infinitely-dilute approximation of the neutron flux. In order to accurately account for spatial and

  13. BWR online monitoring system based on noise analysis

    International Nuclear Information System (INIS)

    Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Alonso, Gustavo; Calleros-Micheland, Gabriel

    2006-01-01

    A monitoring system for during operation early detection of an anomaly and/or faulty behavior of equipment and systems related to the dynamics of a boiling water reactor (BWR) has been developed. The monitoring system is based on the analysis of the 'noise' or fluctuations of a signal from a sensor or measurement device. An efficient prime factor algorithm to compute the fast Fourier transform allows the continuous, real-time comparison of the normalized power spectrum density function of the signal against previously stored reference patterns in a continuously evolving matrix. The monitoring system has been successfully tested offline. Four examples of the application of the monitoring system to the detection and diagnostic of faulty equipment behavior are presented in this work: the detection of two different events of partial blockage at the jet pump inlet nozzle, miss-calibration of a recirculation mass flow sensor, and detection of a faulty data acquisition card. The events occurred at the two BWR Units of the Laguna Verde Nuclear Power Plant. The monitoring system and its possible coupling to the data and processing information system of the Laguna Verde Nuclear Power Plant are described. The signal processing methodology is presented along with the introduction of the application of the evolutionary matrix concept for determining the base signature of reactor equipment or component and the detection of off normal operation conditions

  14. BWR containments license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Smith, S.; Gregor, F.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

  15. 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

  16. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

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

  17. BWR consolidation system final design report

    International Nuclear Information System (INIS)

    Garner, G.L.; Kelly, M.J.; Larsen, W.R.

    1993-05-01

    Because of delays in the opening of a permanent geologic repository to accept spent fuel from nuclear reactor plants, several utilities are seeking additional off-site storage to avert premature shutdown. Fuel rod consolidation is a proven, viable option for pressurized water reactor (PWR) plants, but until now, no consolidation system addressed boiling water reactor (BWR) spent-fuel assemblies.The purpose of this project, jointly funded by the Empire State Electric Energy Research Company (ESEERCO) and the Electric Power Research Institute (EPRI), is to develop a system for consolidating BWR spent fuel assemblies. This design will provide more efficient storage in reactor spent-fuel pools. The design goal is a 2:1 consolidation of the fuel rods and a minimum 10:1 compaction of the non-fuel bearing components. in addition, the consolidation system must be operationally compatible with BWR reactor plants and be economically viable with other forms of supplementary on-site storage. The work began in Lynchburg, Virginia on May 6, 1991 and concluded on September 30, 1992 with the delivery of the final report. The design achieves all of the project goals. Furthermore, consolidation of BWR spent-fuel assemblies is feasible, compatible with reactor plant operations and potentially the lowest cost option for a utility seeking to add oil-site storage capacity

  18. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  19. High-frequency combustion instability control through acoustic modulation at the inlet boundary for liquid rocket engine applications

    Science.gov (United States)

    Bennewitz, John William

    model-predicted mode stability transition was consistent with experimental observations, supporting the premise that inlet acoustic modulation is a means to control high-frequency combustion instabilities. From the modal analysis, it may be deduced that the inlet impedance provides a damping mechanism for instability suppression. Combined, this work demonstrates the strategic application of acoustic modulation within an injector as a potential method to control high-frequency combustion instabilities for liquid rocket engine applications.

  20. Application of the FFTBM method and the power relative contribution to the discharge transitory of the recirculation pumps of a BWR; Aplicacion del metodo FFTBM y de la contribucion relativa de potencia al transitorio de disparo de las bombas de recirculacion de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J.; Fuentes M, L., E-mail: rogelio.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In this work was realized the simulation of the discharge transitory of both recirculation pumps of a BWR with the Simulate-3K code. This type of transitory is used in the stability analyses for the licensing of the fuel reload. An analysis of the precision of the simulation is also presented, using the FFTBM method jointly with the power relative contribution. This way, instead of determining the total precision of the calculation, a weighed precision is obtained by the contribution of each relevant parameter of the transitory. The results show that the precision of the simulation is acceptable due to the small magnitude of the merit figure of the width total average. The error in the merit figure comes mainly from the parameters total flow in the core and temperature of the fuel in the core. (Author)

  1. BWR Servicing and Refueling Improvement Program: Phase I summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

  2. BWR Servicing and Refueling Improvement Program: Phase I summary report

    International Nuclear Information System (INIS)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development

  3. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    Iizuka, N.

    2000-01-01

    The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

  4. TRAB, a transient analysis program for BWR. Part 1

    International Nuclear Information System (INIS)

    Rajamaeki, Markku.

    1980-03-01

    TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)

  5. Shielding analysis of a transport and storage cask for spent BWR fuel applicability of the code SAS4 and discussion of results

    Energy Technology Data Exchange (ETDEWEB)

    Hilbert, F. [Nuclear Cargo and Service GmbH, Hanau (Germany); Morishima, M.; Tamaki, H. [Mitsubishi Heavy Industries, Kobe (Japan)

    2004-07-01

    For the shielding analysis of transport and storage casks for spent fuel the use of computer codes is state of the art. However, in most applications the computer models used for the analysis are simplified to circular geometries to save modelling effort and calculation time. Furthermore, the active zone of the fuel is modelled as homogeneous zone with a uniform average burn-up. In the first part of the present paper it is shown that an exact model is feasible and the effect of the geometrical shape on the dose rates is illustrated. The second part of the paper shows the comparison of the dose rates calculated with 5 different fuel models. Finally, the accuracy of the calculations is discussed.

  6. Kinetic simulations of the Kelvin-Helmholtz instability: Application to boundary layers

    International Nuclear Information System (INIS)

    Thomas, V.A.; Winske, D.

    1994-01-01

    Boundary layers separating the solar wind from planetary magnetic fields and ionospheres are ubiquitous features in the solar system. The transport of plasma between the solar wind and the planetary system is determined by processes that take place at these boundary layers. One such process is the Kelvin-Helmholtz instability, which may be excited due to shear flow at the boundary. Two dimensional hybrid simulations are performed for situations of relevance for this type of boundary layer. For the first situation, the Kelvin-Helmholtz instability is considered within the context of flow past a finite sized obstacle. The finite sized obstacle is a source of plasma and the Kelvin-Helmholtz instability sets in at the interface between the two plasma species. Effects are clearly visible near the subsolar location, indicating that the Kelvin-Helmholtz instability may occur more readily than expected. The source of this unexpected growth can be understood from the overall time dependence of the boundary layer. In the second situation, the Kelvin-Helmholtz instability is studied for the case of a finite sized jet of material moving perpendicular to a magnetic field where the magnetic field reverses sign. Such a scenario may arise as a result of reconnection. The characteristic time and spatial signatures of such an interaction are considered

  7. Kelvin-Helmholtz instabilities and their application to B-type variables

    International Nuclear Information System (INIS)

    Ando, H.

    1981-01-01

    A Kelvin-Helmholtz instability, formed from the differential rotation in the narrow region between the core and envelope, is proposed as a promising mechanism responsible for the excitation of pulsations in B-type variables (53 Per variables and β Cep stars), in which the unstable inertia wave resulting from this instability resonates with an eigenmode of the non-radial oscillation of the whole star. The degeneracy of the two frequencies is found to be expected at any evolutionary stage of a star. The equations for a Kelvin-Helmholtz instability have been formulated for the stellar case, and in the cylindrical configuration limit, the necessary condition for instability and characteristics of the instability have been discussed. It is shown that prograde modes with large /m/ for a given l are excited in almost all cases; which seems to agree with observations. The back reaction of the excited modes on the differential rotation is discussed in these stars, and it is pointed out that the differential rotation can be significantly affected by this effect in a short time. (author)

  8. 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)

  9. Development and Testing of CTF to Support Modeling of BWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gosdin, Chris [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)

    2016-01-29

    This milestone supports developing and assessing COBRA-TF (CTF) for the modeling of boiling water reactors (BWRs). This is achieved in three stages. First, a new preprocessor utility that is capable of handling BWR-specic design elements (e.g., channel boxes and large water rods) is developed. A previous milestone (L3:PHI.CTF.P12.01) led to the development of this preprocessor capability for single assembly models. This current milestone expands this utility so that it is applicable to multi-assembly BWR models that can be modeled in either serial or parallel. The second stage involves making necessary modications to CTF so that it can execute these new models. Specically, this means implementing an outer-iteration loop, specic to BWR models, that equalizes the pressure loss over all assemblies in the core (which are not connected due to the channel boxes) by adjusting inlet mass ow rate. A third stage involves assessing the standard convergence metrics that are used by CTF to determine when a simulation is steady-state. The nal stage has resulted in the implementation of new metrics in the code that give a better indication of how steady the solution is at convergence. This report summarizes these eorts and provides a demonstration of CTF's BWR-modeling capabilities. CASL-U-2016-1030-000

  10. Flex concept for US-A BWR extended loss of AC power events

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Aoyagi, Y.; Kataoka, K. [Toshiba Corporation, Kawasaki, 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 (US-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 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

  11. Flex concept for US-A BWR extended loss of AC power events

    International Nuclear Information System (INIS)

    Powers, J.; Aoyagi, Y.; Kataoka, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

    The US-Advanced Boiling Water Reactor (US-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 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

  12. A numerical study of the nonlinear Rayleigh-Taylor instability, with application of accreting X-ray sources

    International Nuclear Information System (INIS)

    Wang, Y.M.; Nepveu, M.

    1983-01-01

    With a view toward applications to accreting X-ray sources, the Rayleigh-Taylor instability is followed numerically, using a 2-D magnetohydrodynamic code. The presence of a uniform magnetic field in the underlying medium is allowed for. The infalling plasma is found to develop elongated, trailing loops; at least when the initial perturbation is highly symmetric, a narrow neck also forms through the action of the surrounding ram pressure. It is suggested that the swirling motion present in the nonlinear phase could produce some effective large-scale mixing between accreting plasma and the magnetospheric field of a neutron star. Another potentially significant tendency is for the curvature of the infalling plasma pocket to sharpen as the instability develops: magnetic tension may therefore become increasingly effective as a stabilizing influence. (orig.)

  13. Parametric instabilities of rotor-support systems with application to industrial ventilators

    Science.gov (United States)

    Parszewski, Z.; Krodkiemski, T.; Marynowski, K.

    1980-01-01

    Rotor support systems interaction with parametric excitation is considered for both unequal principal shaft stiffness (generators) and offset disc rotors (ventilators). Instability regions and types of instability are computed in the first case, and parametric resonances in the second case. Computed and experimental results are compared for laboratory machine models. A field case study of parametric vibrations in industrial ventilators is reported. Computed parametric resonances are confirmed in field measurements, and some industrial failures are explained. Also the dynamic influence and gyroscopic effect of supporting structures are shown and computed.

  14. Recent developments in BWR water chemistry

    International Nuclear Information System (INIS)

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

  15. BWR mechanics and materials technology update

    International Nuclear Information System (INIS)

    Kiss, E.

    1983-01-01

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

  16. The HAMBO BWR simulator of HAMMLAB

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-15

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

  17. BWR water chemistry guidelines in Japan

    International Nuclear Information System (INIS)

    Hirano, Hideo; Usui, Naoshi; Kitajima, Hideaki

    2014-01-01

    Through Nuclear Power Plant (NPP) operations for more than 40 years in Japan, sustainable developments of water chemistry technologies, including state-of-the-art and unique R and D results, have been developed to achieve the highest performance of maintaining NPP reactor and coolant system component integrity and fuel reliability and excellent dose rate reduction in the world. Based on this long-term operational experience, Boiling Water Reactor (BWR) utilities have established effective and reasonable water chemistry parameters with reference to safety regulations and municipal laws. However, only a few parameters are regulated by control target to ensure the long-term reliability of the reactor and coolant system. After the Fukushima Daiichi Nuclear Power Station accident, it is even more necessary to improve the safety levels of plant operation and increase the transparency of plant management in order to recover the public's confidence forward nuclear energy. Based on these situations, the BWR water chemistry guidelines task group has been working to establish the guidelines under the Standards Committee of Atomic Energy Society of Japan (AESJ). The draft will be deliberated on and approved by the Standards Committee after it receives public comments. The Japanese BWR water chemistry guidelines should provide the world-class strategies for maintaining the long-term integrity of materials and fuel. Concretely, the guidelines should provide reasonable 'control values', 'diagnostic values' and 'action levels' for multiple parameters and suggest the necessary actions to exceed the action levels. This paper provides a framework for the preliminary guidelines for BWR primary water chemistry. (author)

  18. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  19. Decontamination techniques for BWR power generation plant

    International Nuclear Information System (INIS)

    Yoshikawa, Susumu

    1990-01-01

    The present report describes various techniques used for decontamination in BWR power generation plants. Objectives and requirements for decontamination in BWR power plants are first discussed focusing on reduction in dose, prevention of spread of contamination, cleaning of work environments, exposure of equipment parts for inspection, re-use of decontaminated resources, and standards for decontamination. Then, the report outlines major physical, chemical and electrochemical decontamination techniques generally used in BWR power generation plants. The physical techniques include suction of deposits in tanks, jet cleaning, particle blast cleaning, ultrasonic cleaning, coating with special paints, and flushing cleaning. The chemical decontamination techniques include the use of organic acids etc. for dissolution of oxidized surface layers and treatment of secondary wastes such as liquids released from primary decontamination processes. Other techniques are used for removal of penetrated contaminants, and soft and hard cladding in and on equipment and piping that are in direct contact with radioactive materials used in nuclear power generation plants. (N.K.)

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

    International Nuclear Information System (INIS)

    Hernandez S, A.

    2004-01-01

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

  1. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

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

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay 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. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  2. Method of generalized coordinates and an application to Rayleigh-Taylor instability

    International Nuclear Information System (INIS)

    Dienes, J.K.

    1978-01-01

    The method of generalized coordinates is extended to the analysis of continuous bodies for which the degrees of freedom are independent velocity distributions in the spatial coordinates. The corresponding Lagrange equations contain generalized convective terms as well as the usual generalized forces and masses. Since the existence of a potential is not assumed, the equations of motion can be applied to media with arbitrary (possible dissipative) constitutive laws. Material deformation is characterized by the rate of strain, which is taken as the symmetric part of the velocity gradient, making the approach valid for arbitrarily large deformations. As an example, infinitesimal Rayleigh-Taylor instability is considered by analytic methods. Then, large amplitude Rayleigh-Taylor instability is represented with a single-degree-of-freedom analysis that shows the development (by numerical integration) of the known spike-and-bubble configuration of the unstable interface. The infinitesimal stability of a plastically deforming solid and the growth of the instability to large amplitudes are also considered

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  4. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  5. Radiation field control at the latest BWR plants -- design principle, operational experience and future subjects

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shunsuke [Energy Research Lab., Ibaraki (Japan); Ohsumi, Katsumi; Takashima, Yoshie [Hitachi Works, Ibaraki (Japan)

    1995-03-01

    Improvements of operational procedures to control water chemistry, e.g., nickel/iron control, as well as application of hardware improvements for reducing radioactive corrosion products resulted in an extremely low occupational exposure of less than 0.5 man.Sv/yr without any serious impact on the radwaste system, for BWR plants involved in the Japanese Improvement and Standardization Program. Recently, {sup 60}C radioactively in the reactor water has been increasing due to less crud fixation on the two smooth surfaces of new type high performance fuels and to the pH drop caused by chromium oxide anions released from stainless steel structures and pipings. This increase must be limited by changes in water chemistry, e.g., applications of modified nickel/iron ratio control and weak alkali control. Controlled water chemistry to optimize three points, the plant radiation level and integrities of fuel and structural materials, is the primary future subject for BWR water chemistry.

  6. Feasibility study on development of plate-type heat exchanger for BWR plants

    International Nuclear Information System (INIS)

    Ohyama, Nobuhiro; Suda, Kenichi; Ogata, Hiroshi; Matsuda, Shinichi; Nagasaka, Kazuhiro; Fujii, Toshi; Nozawa, Toshiya; Ishihama, Kiyoshi; Higuchi, Tomokazu

    2004-01-01

    In order to apply plate-type heat exchanger to RCW, TCW and FPC system in BWR plants, heat test and seismic test of RCW system heat exchanger sample were carried out. The results of these tests showed new design plate-type heat exchanger satisfied the fixed pressure resistance and seismic resistance and keep the function. The evaluation method of seismic design was constructed and confirmed by the results of tests. As anti-adhesion measure of marine organism, an ozone-water circulation method, chemical-feed method and combination of circulation of hot water and air bubbling are useful in place of the chlorine feeding method. Application of the plate-type heat exchanger to BWR plant is confirmed by these investigations. The basic principles, structure, characteristics, application limit and reliability are stated. (S.Y.)

  7. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

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

    International Nuclear Information System (INIS)

    Palomo, J.; Santiago, J. de

    1988-01-01

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

  9. 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)

  10. Prevention of organic iodide formation in BWR's

    International Nuclear Information System (INIS)

    Karjunen, T.; Laitinen, T.; Piippo, J.; Sirkiae, P.

    1996-01-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR's as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs

  11. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    1999-05-01

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

  12. Streaming sausage, kink and tearing instabilities in a current sheet with applications to the earth's magnetotail

    Science.gov (United States)

    Lee, L. C.; Wang, S.; Wei, C. Q.; Tsurutani, B. T.

    1988-01-01

    This paper investigates the growth rates and eigenmode structures of the streaming sausage, kink, and tearing instabilities in a current sheet with a super-Alfvenic flow. The growth rates and eigenmode structures are first considered in the ideal incompressible limit by using a four-layer model, as well as a more realistic case in which all plasma parameters and the magnetic field vary continuously along the direction perpendicular to the magnetic field and plasma flow. An initial-value method is applied to obtain the growth rate and eigenmode profiles of the fastest growing mode, which is either the sausage mode or kink mode. It is shown that, in the earth's magnetotail, where super-Alfvenic plasma flows are observed in the plasma sheet and the ratio between the plasma and magnetic pressures far away from the current layer is about 0.1-0.3 in the lobes, the streaming sausage and streaming tearing instabilities, but not kink modes, are likely to occur.

  13. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  14. Fuel cycle cost projections. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Clark, L.L.; Chockie, A.D.

    1979-12-01

    This report estimates current and future costs associated with the light water reactor nuclear fuel cycle for both once-through and thermal recycle cases. Using a range of future nuclear power generating scenarios, process flows are developed for each segment of the nuclear fuel cycle. Capital and operating costs are estimated and are combined with the process flows to generate unit cost projections for each fuel cycle segment. The unit costs and process flows are combined in the NUCOST program to estimate fuel cycle power costs through the year 2020. The unit costs are also used to estimate the fuel costs of an individual model PWR and BWR.

  15. Development of membrane moisture separator for BWR off-gas system

    International Nuclear Information System (INIS)

    Ogata, H.; Kawamura, S.; Kumasaka, M.; Nishikubo, M.

    2001-01-01

    In BWR plant off-gas treatment systems, dehumidifiers are used to maintain noble gas adsorption efficiency in the first half of the charcoal hold-up units. From the perspective of simplifying and reducing the cost of such a dehumidification system, Japanese BWR utilities and plant fabricators have been developing a dehumidification system employing moisture separation membrane of the type already proven in fields such as medical instrumentation and precision measuring apparatus. The first part of this development involved laboratory testing to simulate the conditions found in an actual off-gas system, the results of which demonstrated satisfactory results in terms of moisture separation capability and membrane durability, and suggested favorable prospects for application in actual off-gas systems. Further, in-plant testing to verify moisture separation capability and membrane durability in the presence of actual gases is currently underway, with results so far suggesting that the system is capable of obtaining good moisture separation capability. (author)

  16. 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)

  17. Phenomenology of BWR fuel assembly degradation

    Science.gov (United States)

    Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

    2018-03-01

    Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

  18. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  19. 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

  20. 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.

  1. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

    2013-01-01

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  2. Dose rate reduction method for NMCA applied BWR plants

    International Nuclear Information System (INIS)

    Nagase, Makoto; Aizawa, Motohiro; Ito, Tsuyoshi; Hosokawa, Hideyuki; Varela, Juan; Caine, Thomas

    2012-09-01

    results showed that the Co-60 amount on the coupon with Pt was about 40 % smaller than on the coupon without Pt at 500 hours under no hydrogen peroxide addition case. Hydrogen peroxide had a tendency to accelerate the Co-60 deposition even under low concentration of below 5 ppb. Nevertheless, the presence of Pt on a surface caused a reduction of about 50% in the Co-60 deposition. The above results suggest that recontamination after chemical decontamination is reduced by the presence Pt on these surfaces. This idea is consistent with in-plant tests. In these tests coupons pre-filmed under the NWC conditions were set in a plant's MMS (Mitigation Monitoring System) along with coupons that had been treated by the HOP method. Both of these coupons were inserted on the MMS just before OLNC applications were performed in two BWR plants; Cooper and River Bend. These coupons were removed from the MMS after about 7 months from the installation. Co-60 on the coupon surfaces were dissolved and quantified. Though the absolute value of Co-60 amount was not equal in the coupons from the two plants, in both sets of coupons the amount of Co-60 on the decontaminated coupons was about half of the Co-60 measured on the pre-filmed coupons. These results indicate that the presence of platinum before an oxide layer is generated reduces the amount of Co-60 incorporation. The Co deposition coefficient calculated from the Co-60 amounts of the decontaminated coupons, Co-60 concentration in the reactor water and the deposition term was very small compared to that calculated from many plants data. The combination of chemical decontamination followed by the Pt deposition before operation and Zn injection under HWC condition was considered to be effective method to accomplish the low BRAC dose rate in NMCA applied BWR plants based on the laboratory experiments and in-plant tests. (authors)

  3. The Parametric Decay Instability of Alfvén Waves in Turbulent Plasmas and the Applications in the Solar Wind

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Mijie; Xiao, Chijie; Wang, Xiaogang [State Key Laboratory of Nuclear Physics and Technology, Fusion Simulation Center, School of Physics, Peking University, Beijing 100871 (China); Li, Hui, E-mail: cjxiao@pku.edu.cn [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2017-06-10

    We perform three-dimensional ideal magnetohydrodynamic (MHD) simulations to study the parametric decay instability (PDI) of Alfvén waves in turbulent plasmas and explore its possible applications in the solar wind. We find that, over a broad range of parameters in background turbulence amplitudes, the PDI of an Alfvén wave with various amplitudes can still occur, though its growth rate in turbulent plasmas tends to be lower than both the theoretical linear theory prediction and that in the non-turbulent situations. Spatial–temporal FFT analyses of density fluctuations produced by the PDI match well with the dispersion relation of the slow MHD waves. This result may provide an explanation of the generation mechanism of slow waves in the solar wind observed at 1 au. It further highlights the need to explore the effects of density variations in modifying the turbulence properties as well as in heating the solar wind plasmas.

  4. Thermal-hydraulics stability of natural circulation BWR under startup. Flashing effects

    International Nuclear Information System (INIS)

    Hu, Rui; Kazimi, Mujid S.

    2009-01-01

    To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a boiling natural circulation reactor like the ESBWR. In such systems, thermal-hydraulic stability during low pressure start-up should be examined while considering the flashing induced by the pressure drop in the channel and the chimney due to gravity head. In this work, a BWR stability analysis code in the frequency domain, named FISTAB (Flashing-Induced STability Analysis for BWR), was developed to address the issue of flashing-induced instability. A thermal-hydraulics non-homogeneous equilibrium model (NHEM) based on a drift flux formulation along with a lumped fuel dynamics model is incorporated in the work. The vapor generation rate is derived from the mixture energy conservation equation while considering the effect of flashing. The functionality of the FISTAB code was confirmed by comparison to experimental results from SIRIUS-N facility at CRIEPI, Japan. Both stationary and perturbation results agree well with the experimental results. (author)

  5. New Density Estimation Methods for Charged Particle Beams With Applications to Microbunching Instability

    Energy Technology Data Exchange (ETDEWEB)

    Balsa Terzic, Gabriele Bassi

    2011-07-01

    In this paper we discuss representations of charge particle densities in particle-in-cell (PIC) simulations, analyze the sources and profiles of the intrinsic numerical noise, and present efficient methods for their removal. We devise two alternative estimation methods for charged particle distribution which represent significant improvement over the Monte Carlo cosine expansion used in the 2d code of Bassi, designed to simulate coherent synchrotron radiation (CSR) in charged particle beams. The improvement is achieved by employing an alternative beam density estimation to the Monte Carlo cosine expansion. The representation is first binned onto a finite grid, after which two grid-based methods are employed to approximate particle distributions: (i) truncated fast cosine transform (TFCT); and (ii) thresholded wavelet transform (TWT). We demonstrate that these alternative methods represent a staggering upgrade over the original Monte Carlo cosine expansion in terms of efficiency, while the TWT approximation also provides an appreciable improvement in accuracy. The improvement in accuracy comes from a judicious removal of the numerical noise enabled by the wavelet formulation. The TWT method is then integrated into Bassi's CSR code, and benchmarked against the original version. We show that the new density estimation method provides a superior performance in terms of efficiency and spatial resolution, thus enabling high-fidelity simulations of CSR effects, including microbunching instability.

  6. Evaluation of BWR emergency procedure guidelines for BWR ATWS using RAMONA-3B code

    International Nuclear Information System (INIS)

    Neymotin, L.; Slovik, G.; Cazzoli, E.; Saha, P.

    1985-01-01

    An MSIV Closure ATWS calculation for a typical BWR/4 (Browns Ferry, Unit 1) was performed using the RAMONA-3B code which is a BWR systems transient code combining three-dimensional neutronic core representation with multi-channel one-dimensional thermal hydraulics. The main objective of the study was to perform a best-estimate evaluation of the recently proposed Emergency Procedure Guidelines for Anticipated Transients Without Scram (ATWS). Emphasis was placed on evaluating the effects of lowering the downcomer water level to the Top of Active Fuel (TAF) and vessel depressurization. The calculation was run up to approximately 1200 seconds. Both actions, namely, lowering the water level and vessel depressurization, lowered the reactor power to some extent. However, the pressure suppression pool water temperature still reached approximately 90 0 C (potential High Pressure Coolant Injection (HPCI) pump seal failure temperature) in twenty minutes. Thus, other actions such as boron injection and/or manual control rod insertion are necessary to mitigate a BWR/4 Main Steam Isolation Valve (MSIV) closure ATWS. 19 refs., 14 figs., 3 tabs

  7. BWR stability: history and state-of-the-art

    International Nuclear Information System (INIS)

    Yadigaroglu, George

    2014-01-01

    The paper briefly recalls the historical developments, reviews the important phenomena, the analytical and simulation tools that are used for the analysis of BWR stability focussing on the linear, frequency domain methods

  8. Thermal-hydraulic stability tests for newly designed BWR rod bundle (step-III fuel type A)

    International Nuclear Information System (INIS)

    Mitsutake, T.; Chuman, K.; Miura, S.; Morooka, S.; Moriya, K.; Kitamura, H.; Toba, A.; Omoto, A.

    2004-01-01

    Thermal-hydraulic stability tests have been performed on electrically heated bundles to simulate the newly designed Boiling Water Reactor (BWR) fuels in a parallel channel test loop. The objective of the current experimental program is to investigate how the newly designed bundle could improve the thermal-hydraulic stability. Measurements of the thermal-hydraulic instability thresholds in two vertical rod bundles have been conducted in steam-water two-phase flow conditions at the TOSHIBA test loop. Fluid conditions were BWR operating conditions of 7 MPa system pressure, 1.0-2.0x10 6 kg/m 2 /h inlet mass flux and 28-108 kJ/kg inlet subcooling. The parallel channel test loop consists of a main bundle of 3x3 indirectly heated rods of 1/9 symmetry of 9x9 full lattice and a bypass bundle of 8x8. These are both simulated BWR rod bundles in respect of rod diameter, heated length, rod configuration, fuel rod spacer, core inlet hydraulic resistance and upper tie plate. There are three types of the 3x3 test bundles with different configurations of a part length rod of two-thirds the length of the other rods and an axial power shape. The design innovation of the part length rod for a 9x9 lattice development, though addition of more fuel rods increases bundle pressure drop, reduces pressure drop in the two-phase portion of the bundle, and enhances the thermal hydraulic stability. Through the experiments, the parameter dependency on the channel stability threshold is obtained for inlet subcooling, inlet mass flux, inlet flow resistance, axial power shape and part length rod. The main conclusion is that the stability threshold is about 10% greater with the part length rod than without the part length rod. The new BWR bundle consisting of the part length rod has been verified in respect of thermal hydraulic stability performance. (author)

  9. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Kato, Shigeru

    1996-01-01

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  10. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  11. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  12. Fast spectrum transmutation in a BWR

    International Nuclear Information System (INIS)

    Wallenius, J.; Westlen, D.

    2007-01-01

    We propose an innovative fuel design for boiling water reactors, where the use of metallic alloy fuel and hafnium clad results in a fast neutron spectrum, suitable for transmutation of minor actinides. Monte Carlo calculations made with the JEFF3.1 data library show that in the top of an up-rated BWR, it is possible to achieve fission probabilities for even neutron number nuclides similar to those of sodium-cooled reactors. Thus, from a strict neutronic perspective, multi-recycling of americium and curium could be performed in the top of BWRs. Fuel and coolant temperature feedbacks remain within acceptable ranges, but control rod worths are reduced to the extent that enriched boron might have to be used to ensure sufficient shutdown margins. (authors)

  13. 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

  14. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  15. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. Agitation device in BWR type reactor

    International Nuclear Information System (INIS)

    Nakamizo, Hiroshi.

    1988-01-01

    Purpose: To secure the integrity of a pressure vessel upon reactor accident by maintaining the water temperature in the pressure suppression pool at a low temperature. Constitution: The pressure vessel and the pressure control pool in a BWR type reactor are connected with a vent pipe by way of a safety valve. Steams generated in the pressure vessel are introduced through the vent pipe into pool water. A turbine is disposed at the midway of the vent pipe and rotated with steams flowing through the vent pipe. A vane is disposed in the pressure suppression pool for agitating water and rotated by the turbine. The temperature distribution for the water in the pool is made uniform by the stirring. In this way, by rotating the vane using steams, since water can automatically be agitated as required without utilizing any power source of other system such as an electric power supply, pool water can be cooled at a high reliability. (Ikeda, J.)

  17. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  18. Adaptative control of aero-acoustic instabilities. Application to propulsion systems; Controle adaptatif des instabilites aeroacoustiques. Application aux systemes de propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Mettenleiter, M.

    2000-02-15

    This work treats active adaptive control of aero-acoustic instabilities. In particular, we are interested in an application to solid propellant rockets. The study is part of the research program ASSM coordinated by CNES and ONERA and the aim is to increase the performance of the P230 segmented solid propellant boosters of the Ariane 5 rocket. The work has been carried out in collaboration with other partners of this program. The main objective of this study is the development of control algorithms, able to diminish low frequency instabilities encountered in propulsion systems. First, the instability phenomenon is analyzed in a simplified experimental setup and similarity is shown with instabilities observed in real propulsion systems. This study enables us to conceive adaptive control strategies, which have been tested on three different levels: - In a simplified dynamical simulation; - During an experimental study; - Using full numerical simulations. The three levels of application made it possible to study the behaviour of the different control strategies. We could show that the actuator signal modifies the behaviour of the system on the acoustic level. But as there is a strong interaction between the pressure fluctuations and the hydrodynamic behaviour, the flow structure is also modified by active control. This behaviour corresponds to the simplified model of the phenomenon, which has been used to define the control algorithms. The control action 'at the noise source' makes it possible to distinguish this kind of algorithms from schemes based on the anti-noise principle. After this first part, where we showed the feasibility of control, we particularly considered algorithms which can act in an unknown environment. The information about the system behaviour. which is necessary for convergence of the controller is now obtained in parallel during control. An identification off-line, used at the beginning of the research, is no longer necessary. Self

  19. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    Ortiz S, J.J.

    1998-01-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  20. Carpal instability

    International Nuclear Information System (INIS)

    Schmitt, R.; Froehner, S.; Coblenz, G.; Christopoulos, G.

    2006-01-01

    This review addresses the pathoanatomical basics as well as the clinical and radiological presentation of instability patterns of the wrist. Carpal instability mostly follows an injury; however, other diseases, like CPPD arthropathy, can be associated. Instability occurs either if the carpus is unable to sustain physiologic loads (''dyskinetics'') or suffers from abnormal motion of its bones during movement (''dyskinematics''). In the classification of carpal instability, dissociative subcategories (located within proximal carpal row) are differentiated from non-dissociative subcategories (present between the carpal rows) and combined patterns. It is essential to note that the unstable wrist initially does not cause relevant signs in standard radiograms, therefore being ''occult'' for the radiologic assessment. This paper emphasizes the high utility of kinematographic studies, contrast-enhanced magnetic resonance imaging (MRI) and MR arthrography for detecting these predynamic and dynamic instability stages. Later in the natural history of carpal instability, static malalignment of the wrist and osteoarthritis will develop, both being associated with significant morbidity and disability. To prevent individual and socio-economic implications, the handsurgeon or orthopedist, as well as the radiologist, is challenged for early and precise diagnosis. (orig.)

  1. CFD predictions of standby liquid control system mixing in lower plenum of a BWR

    International Nuclear Information System (INIS)

    Boyd, Christopher; Skarda, Raymond

    2014-01-01

    Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

  2. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

    2006-01-01

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  3. Development status of compact containment BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Mori, H.; Sekiguchi, K.; Kuroki, M.; Arai, K.; Hida, T.

    2005-01-01

    In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. 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, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response

  4. BWR Source Term Generation and Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.C. Ryman

    2003-07-31

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-00006 REV 01. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operating conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the

  5. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  6. Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island Design (Docket No. 50-447). Supplement No. 3

    International Nuclear Information System (INIS)

    1985-01-01

    Supplement 3 to the Safety Evaluation Report (SER) for the application filed by General Electric Company for the final design approval for the GE BWR/6 nuclear island design has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report supplements the GESSAR II SER (NUREG-0979), issued in April 1983, summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design. Subject to favorable resolution of the items discussed in this supplement, the staff concludes that the GESSAR II design satisfactorily addresses the severe-accident concerns described in draft NUREG-1070

  7. Recent technology for BWR operator training simulators

    International Nuclear Information System (INIS)

    Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

    1990-01-01

    As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)

  8. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  9. Method of operating BWR type power plants

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To improve the operation efficiency of BWR type reactors by reducing the time from the start-up of the reactor to the start-up of the turbine and electrical generator, as well as decrease the pressure difference in each of the sections of the pressure vessel to thereby extend its life span. Method: The operation comprises switching the nuclear reactor from the shutdown mode to the start-up mode, increasing the reactor power to a predetermined level lower than a rated power while maintaining the reactor pressure to a predetermined level lower than a rated pressure, starting up a turbine and an electrical generator in the state of the predetermined reactor pressure and the reactor power to connect the electrical generator to the power transmission system and, thereafter, increasing the reactor pressure and the reactor power to the predetermined rated pressure and rated power respectively. This can shorten the time from the start-up of the reactor to the start of the power transmission system, whereby the operation efficiency of the power plant can be improved. (Moriyama, K.)

  10. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  11. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  12. On the possibility of a decay ratio jump under continuous parameter variation in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lange, Carsten; Hennig, Dieter; Hurtado, Antonio [Technische Univ. Dresden (Germany). Professur fuer Wasserstoff- und Kernenergietechnik

    2012-11-01

    The objective of this paper is to demonstrate that specific methods of the nonlinear dynamics like bifurcation analysis are helpful to understand the temporal behavior of the nonlinear dynamical system BWR in many details. So, the stability state of the system changes discontinuously if a nonlinear dynamical system encountered a Hopf bifurcation under parameter variation. BWR stability analysis is performed in the most cases by application of system codes which provide the time evolution of the neutron flux or thermal power at a defined operational point (OP) after imposing a system parameter perturbation (like a control rod sinusoidal movement). However, in general, we are not able to understand the real stability state of the BWR at a specific OP by application of system code analysis alone. We have shown in the paper that complex dynamical states with coexisting stability modes (fixed points, stable and unstable limit cycles) could spontaneously emerge under selected system parameter variations. Hence, even in the relative simple case where stable fixed points and unstable limit cycles coexist, in the framework of a stability test the unstable dynamical mode could be overlooked because the simple stability linear indicator ''decay ratio'' does not indicate an unstable state for small amplitude perturbations. Or, e.g., both, the in-phase and the out-of-phase oscillation mode can encounter a Hopf bifurcation (everyone for itself) resulting in a change of the stability characteristics which cannot be ''detected'' by the asymptotic decay ratio. (orig.)

  13. 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)

  14. Advances in the discontinuous Galerkin method: Hybrid schemes and applications to the reactive infiltration instability in an upwelling compacting mantle

    Science.gov (United States)

    Schiemenz, Alan R.

    High-order methods are emerging in the scientific computing community as superior alternatives to the classical finite difference, finite volume, and continuous finite element methods. The discontinuous Galerkin (DG) method in particular combines many of the positive features of all of these methods. This thesis presents two projects involving the DG method. First, a Hybrid scheme is presented, which implements DG areas where the solution is considered smooth, while dropping the order of the scheme elsewhere and implementing a finite volume scheme with high-order, non-oscillatory solution reconstructions suitable for unstructured mesh. Two such reconstructions from the ENO class are considered in the Hybrid. Successful numerical results are presented for nonlinear systems of conservation laws in one dimension. Second, the high-order discontinuous Galerkin and Fourier spectral methods are applied to an application modeling three-phase fluid flow through a porous medium, undergoing solid-fluid reaction due to the reactive infiltration instability (RII). This model incorporates a solid upwelling term and an equation to track the abundance of the reacting mineral orthopyroxene (opx). After validating the numerical discretization, results are given that provide new insight into the formation of melt channels in the Earth's mantle. Mantle heterogeneities are observed to be one catalyst for the development of melt channels, and the dissolution of opx produces interesting bifurcations in the melt channels. An alternative formulation is considered where the mass transfer rate relative to velocity is taken to be infinitely large. In this setting, the stiffest terms are removed, greatly reducing the cost of time integration.

  15. Development of a detailed BWR core thermal-hydraulic analysis method based on the Japanese post-BT standard using a best-estimate code

    International Nuclear Information System (INIS)

    Ono, H.; Mototani, A.; Kawamura, S.; Abe, N.; Takeuchi, Y.

    2004-01-01

    The post-BT standard is a new fuel integrity standard or the Atomic Energy Society of Japan that allows temporary boiling transition condition in the evaluation for BWR anticipated operational occurrences. For application of the post-BT standard to BWR anticipated operational occurrences evaluation, it is important to identify which fuel assemblies and which axial, radial positions of fuel rods have temporarily experienced the post-BT condition and to evaluates how high the fuel cladding temperature rise was and how long the dryout duration continued. Therefore, whole bundle simulation, in which each fuel assembly is simulated independently by one thermal-hydraulic component, is considered to be an effective analytical method. In the present study, a best-estimate thermal-hydraulic code, TRACG02, has been modified to extend it predictive capability by implementing the post-BT evaluation model such as the post-BT heat transfer correlation and rewetting correlation and enlarging the number of components used for BWR plant simulation. Based on new evaluation methods, BWR core thermal-hydraulic behavior has been analyzed for typical anticipated operational occurrence conditions. The location where boiling transition occurs and the severity of fuel assembly in the case of boiling transition conditions such as fuel cladding temperature, which are important factors in determining whether the reuse of the fuel assembly can be permitted, were well predicted by the proposed evaluation method. In summary, a new evaluation method for a detailed BWR core thermal-hydraulic analysis based on the post-BT standard of the Atomic Energy Society of Japan has been developed and applied to the evaluation of the post-BT standard during the actual BWR plant anticipated operational occurrences. (author)

  16. Fuel relocation effects in BWR LOCA conditions

    International Nuclear Information System (INIS)

    Raul Orive Moreno; Ines Gallego Cabezon; Pablo Julio Garcia Sedano; Yolanda Tofino Gomez; Pedro Mata Alonso

    2005-01-01

    One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MT rod average burnup. One of the licensing points of concern is the behavior of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/MT and H.B. Robinson at 67 GWd/MT. ANL LOCA tests indicate potential fuel relocation during LOCA. This could result in an increase of LHGR during a real plant LOCA. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations with GE-14 fuel design for the whole exposure range, quantifying fuel relocation impact. This effect has been modeled and implemented in FRAP-T6/APK (vendor independent IBERDROLA licensing thermomechanical code), as well as the wall-to-fluid heat transfer area increase in the ballooned region. Separate and combined impacts on PCT and ECR values can be evaluated with this modified code version. A new hoop strain versus rupture temperature curve is implemented in code, starting from NUREG-0630 model data base, but with a more best-estimate fit, in order to reproduce expected experimental values. The increase of heat transfer in the ballooned region has been validated with Halden LOCA tests. Preliminary results indicate that the effect of fuel relocation is expected to be compensated by the increased heat transfer area. This effect is to be confirmed with the Halden LOCA tests in progress. (author)

  17. Simulation of Irradiated BWR fuel rod (TS) test in NSRR using FRAP-T6 and NSR-77

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Murofushi, Akira; Hosoyamada, Ryuji.

    1994-03-01

    Series of pulse irradiation tests have been performed in the Nuclear Safety Research Reactor (NSRR) to investigate irradiated fuel rod performance under the Reactivity Initiated Accident (RIA) conditions. Five tests, called Tests TS-1 through TS-5, were conducted in a period from 1989 to 1993 with irradiated 7x7 type BWR fuel rods provided from a commercial power plant. Simulation calculations of the TS tests were carried out with the FRAP-T6 code, which is widely used in the world to estimate fuel performance under various accident conditions, and with the NSR77 code, which describes fresh fuel rod performance well in the NSRR tests. Results of the calculation are compiled in this report and applicability of the codes to the irradiated BWR fuel rod tests is discussed. (author)

  18. Condensate treatment in BWR circuits by filter demineralizer units using powdered ion exchange resin at medium and high temperature

    International Nuclear Information System (INIS)

    De Martino, R.

    1983-01-01

    Considering the radiation build-up in some BWR reactors, we make a correlation between this phenomenon and the condensate purification system applied and the point of its utilization into the circuits. The application temperature of such a plant seems to have a very important role on the equilibria of metals contained in the reactor water and on the oxide composition. The efficiency of the condensate polishing system and the corrosion control are the most interesting objectives to achieve and to maintain, to control and regulate the physical and chemical process in the feedwater and in the reactor water. Up to date the technology owns major knowledge and a consistent know-how on using chemical products in order to increase the condensate polishing system efficiency. It is also considered a typical parallel case of a conventional power station and a secondary system of BWR units. (author)

  19. A general approach to optomechanical parametric instabilities

    International Nuclear Information System (INIS)

    Evans, M.; Barsotti, L.; Fritschel, P.

    2010-01-01

    We present a simple feedback description of parametric instabilities which can be applied to a variety of optical systems. Parametric instabilities are of particular interest to the field of gravitational-wave interferometry where high mechanical quality factors and a large amount of stored optical power have the potential for instability. In our use of Advanced LIGO as an example application, we find that parametric instabilities, if left unaddressed, present a potential threat to the stability of high-power operation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-07-01

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

  1. Comparison of the General Electric BWR/6 standard plant design to the IAEA NUSS codes and guides

    International Nuclear Information System (INIS)

    D'Ardenne, W.H.; Sherwood, G.G.

    1985-01-01

    The General Electric BWR/6 Mark III standard plant design meets or exceeds current requirements of published International Atomic Energy Agency (IAEA) Nuclear Safety Standards (NUSS) codes and guides. This conclusion is based on a review of the NUSS codes and guides by General Electric and by the co-ordinated US review of the NUSS codes and guides during their development. General Electric compared the published IAEA NUSS codes and guides with the General Electric design. The applicability of each code and guide to the BWR/6 Mark III standard plant design was determined. Each code or guide was reviewed by a General Electric engineer knowledgeable about the structures, systems and components addressed and the technical area covered by that code or guide. The results of this review show that the BWR/6 Mark III standard plant design meets or exceeds the applicable requirements of the published IAEA NUSS codes and guides. The co-ordinated US review of the IAEA NUSS codes and guides corroborates the General Electric review. In the co-ordinated US review, the USNRC and US industry organizations (including General Electric) review the NUSS codes and guides during their development. This review ensures that the NUSS codes and guides are consistent with the current US government regulations, guidance and regulatory practices, US voluntary industry codes and standards, and accepted US industry design, construction and operational practices. If any inconsistencies are identified, comments are submitted to the IAEA by the USNRC. All US concerns submitted to the IAEA have been resolved. General Electric design reviews and the Final Design Approval (FDA) issued by the USNRC have verified that the General Electric BWR/6 Mark III standard plant design meets or exceeds the current US requirements, guidance and practices. Since these requirements, guidance and practices meet or exceed those of the NUSS codes and guides, so does the General Electric design. (author)

  2. Decay heat measurements and predictions of BWR spent fuel

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Heeb, C.M.; Creer, A.M.

    1986-06-01

    Pre-calorimetry decay heat predictions obtained with the ORIGEN2 computer code were compared to calorimeter data obtained for eleven boiling water reactor (BWR) spent fuel assemblies using General Electric, Morris Operation's in-pool calorimeter. Ten of the 7 x 7 BWR spent fuel assemblies were obtained from Nebraska Public Power District's Cooper Nuclear Station. The remaining BWR assembly was from Commonwealth Edison's Dresden Nuclear Power Plant. The assemblies had burnups ranging from 5.3 to 27.6 GWD/MTU and had been removed from their respective reactors for 2 or more years. The majority of the assemblies had burnups of between 20 and 28 GWD/MTU and had been out of the reactor 2 to 4 years. The assemblies represent spent fuel that has been continuously burned and fuel that has been reinserted. Comparisons of ORIGEN2 pre-calorimetry decay heat predictions with calorimeter data showed that predictions agreed with data within the precision/repeatibility of the experimental data (+-15 Watts or 5% for a 300 Watt BWR assembly). Comparisons of predicted axial gamma profiles based on core-averaged axial burnups with measured profiles showed difference. These differences may be explained by reactor operation with partially inserted control rods

  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. Operator training simulator for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Watanabe, Tadasu

    1988-01-01

    For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

  5. FINANCIAL INSTABILITY AND POLITICAL INSTABILITY

    Directory of Open Access Journals (Sweden)

    Ionescu Cristian

    2012-12-01

    Full Text Available There is an important link between the following two variables: financial instability and political instability. Often, the link is bidirectional, so both may influence each other. This is way the lately crisis are becoming larger and increasingly complex. Therefore, the academic environment is simultaneously talking about economic crises, financial crises, political crises, social crises, highlighting the correlation and causality between variables belonging to the economic, financial, political and social areas, with repercussions and spillover effects that extend from one area to another. Given the importance, relevance and the actuality of the ones described above, I consider that at least a theoretical analysis between economic, financial and political factors is needed in order to understand the reality. Thus, this paper aims to find links and connections to complete the picture of the economic reality.

  6. A Method for Avoiding Numerical Instability in FDTD-based Surge Simulations and its Application to Representation of Thin Wires

    Science.gov (United States)

    Tatematsu, Akiyoshi; Noda, Taku

    The FDTD (Finite Difference Time Domain) method, which is used for the computation of electromagnetic fields, is a useful tool for surge problems in three-dimensional arrangements, which cannot be rigorously treated by simulation methods such as EMTP (Electromagnetic Transients Programs) based on circuit theory. In the FDTD method, thin wires such as power lines are often represented by forcing the electric fields along the wires to zero. If a thin wire is represented in this way, its radius is determined by the size of the cells surrounding the wire. Although the technique for simulating an arbitrary radius has already been proposed, numerical instability often occurs in simulations when the wire radius is much smaller than the size of the cells surrounding the wire. As a conventional countermeasure against such numerical instability, the time discretization is often set to a smaller value than that based on Courant's condition. However, the smaller time discretization results in an increase in the calculation time. We propose a new technique for avoiding the numerical instability. Using this technique, in only a part of an analysis space where the numerical instability may occur, the time discretization is set to a value smaller than that based on Courant's condition, while the time discretization is set to a value close to that based on Courant's condition in the other part of the analysis space. We performed an FDTD-based simulation of a thin wire using the proposed numerical stabilization technique for validation.

  7. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  8. An economic analysis of BWR control rod blade management strategies. Final report

    International Nuclear Information System (INIS)

    Welsh, J.

    1995-12-01

    Nuclear power plants have available a number of alternative courses of action that can contribute to the reduction of personnel exposure to radiation. Possible actions at boiling water reactor (BWR) plants include accelerating the replacement of high-cobalt control rod blades (CRB) or the blades' high-cobalt pins and rollers with low or non-cobalt substitutes. To help utilities understand the exposure reduction and the economic costs and benefits associated with management alternatives, such as accelerated replacement of blades, pins and rollers, EPRI has initiated a project called Cost/Benefit Software for Analyses of Radiation Control Measures (RP1935-32). Through this project EPRI will incorporate engineering-economic techniques into a series of analytical tools that will provide useful insights about alternative exposure reduction options. Prototype software has been developed in an Excel worksheet to analyze issues associated with BWR control rod blade management options. The CRB replacement problem framework and analysis methodology incorporated into the software tool will help plant managers consider explicitly key engineering and economic issues that are relevant to exposure reduction decisions. This tool generates results that can help plant managers make decisions that are fiscally wise by showing all the cost and benefit implications associated with a management action under consideration. This report describes the general analytical approach for evaluating exposure reduction alternatives. The methodology used to analyze blade and pin and roller replacement alternatives, and the results of a case study application of the methodology and the software prototype at Commonwealth Edison

  9. Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

    International Nuclear Information System (INIS)

    Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

    2017-01-01

    BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

  10. A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors

    International Nuclear Information System (INIS)

    Alvarez G, G.

    1991-01-01

    The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

  11. Quantitative evaluation for training results of nuclear plant operator on BWR simulator

    International Nuclear Information System (INIS)

    Sato, Takao; Sato, Tatsuaki; Onishi, Hiroshi; Miyakita, Kohji; Mizuno, Toshiyuki

    1985-01-01

    Recently, the reliability of neclear power plants has largely risen, and the abnormal phenomena in the actual plants are rarely encountered. Therefore, the training using simulators becomes more and more important. In BWR Operator Training Center Corp., the training of the operators of BWR power plants has been continued for about ten years using a simulator having the nearly same function as the actual plants. The recent high capacity ratio of nuclear power plants has been mostly supported by excellent operators trained in this way. Taking the opportunity of the start of operation of No.2 simulator, effort has been exerted to quantitatively grasp the effect of training and to heighten the quality of training. The outline of seven training courses is shown. The technical ability required for operators, the items of quantifying the effect of training, that is, operational errors and the time required for operation, the method of quantifying, the method of collecting the data and the results of the application to the actual training are described. It was found that this method is suitable to quantify the effect of training. (Kako, I.)

  12. Design and axial optimization of nuclear fuel for BWR reactors

    International Nuclear Information System (INIS)

    Garcia V, M.A.

    2006-01-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  13. Evaluation of application of a carpal brace as a treatment for carpal ligament instability in dogs: 14 cases (2008-2011).

    Science.gov (United States)

    Tomlinson, Julia E; Manfredi, Jane M

    2014-02-15

    To determine whether carpal brace application is a viable treatment for dogs with unilateral carpal ligament instability. Retrospective case series. 14 client-owned athletic dogs. Medical records were reviewed to identify dogs treated with a brace for unilateral carpal valgus or varus instability between August 2008 and August 2011. Treatment included passive motion and isometric strengthening exercises during brace application. Of the 14 dogs, 11 were considered to have returned to normal function; 11 of 12 dogs returned to agility competition. Carpal measurements before treatment indicated the affected limb had significantly greater valgus measurements (median, 30°; range, 30° to 35°), significantly greater varus measurements (median, 15°; range, 15° to 25°), and significantly less flexion (median, 37.5°; range, 30° to 45°), compared with results for the contralateral carpus. Long-term monitoring revealed no differences in measurements between affected and contralateral limbs. Valgus measurements of the affected carpus at brace removal (median, 15°; range, 15° to 20°) and at the end of long-term monitoring (median, 15°; range, 15° to 20°) were significantly lower than measurements before treatment (median, 30°; range, 30° to 35°). Dogs had significantly lower lameness scores (assessed on a scale of 0 to 5) at brace removal (median, 0; range, 0) and at the end of monitoring (median, 0; range, 0 to 2), compared with scores before treatment (median, 3; range, 1 to 3). Application of a carpal brace resulted in improved stability and resolution or reduction in lameness in dogs with carpal ligament instability.

  14. Developpement of a numerical method for Navier-Stokes equations in anelastic approximation: application to Rayleigh-Taylor instabilities

    International Nuclear Information System (INIS)

    Hammouch, Z.

    2012-01-01

    The 'anelastic' approximation allows us to filter the acoustic waves thanks to an asymptotic development of the Navier-Stokes equations, so increasing the averaged time step, during the numerical simulation of hydrodynamic instabilities development. So, the anelastic equations for a two fluid mixture in case of Rayleigh-Taylor instability are established.The linear stability of Rayleigh-Taylor flow is studied, for the first time, for perfect fluids in the anelastic approximation. We define the Stokes problem resulting from Navier-Stokes equations without the non linear terms (a part of the buoyancy is considered); the ellipticity is demonstrated, the Eigenmodes and the invariance related to the pressure are detailed. The Uzawa's method is extended to the anelastic approximation and shows the decoupling speeds in 3D, the particular case k = 0 and the spurious modes of pressure. Passing to multi-domain allowed to establish the transmission conditions.The algorithms and the implementation in the existing program are validated by comparing the Uzawa's operator in Fortran and Mathematica languages, to an experiment with incompressible fluids and results from anelastic and compressible numerical simulations. The study of the influence of the initial stratification of both fluids on the development of the Rayleigh-Taylor instability is initiated. (author) [fr

  15. Instability timescale for the inclination instability in the solar system

    Science.gov (United States)

    Zderic, Alexander; Madigan, Ann-Marie; Fleisig, Jacob

    2018-04-01

    The gravitational influence of small bodies is often neglected in the study of solar system dynamics. However, this is not always an appropriate assumption. For example, mutual secular torques between low mass particles on eccentric orbits can result in a self-gravity instability (`inclination instability'; Madigan & McCourt 2016). During the instability, inclinations increase exponentially, eccentricities decrease (detachment), and orbits cluster in argument of perihelion. In the solar system, the orbits of the most distant objects show all three of these characteristics (high inclination: Volk & Malhotra (2017), detachment: Delsanti & Jewitt (2006), and argument of perihelion clustering: Trujillo & Sheppard (2014)). The inclination instability is a natural explanation for these phenomena.Unfortunately, full N-body simulations of the solar system are unfeasible (N ≈ O(1012)), and the behavior of the instability depends on N, prohibiting the direct application of lower N simulations. Here we present the instability timescale's functional dependence on N, allowing us to extrapolate our simulation results to that appropriate for the solar system. We show that ~5 MEarth of small icy bodies in the Sedna region is sufficient for the inclination instability to occur in the outer solar system.

  16. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  17. Distributed control and data processing system with a centralized database for a BWR power plant

    International Nuclear Information System (INIS)

    Fujii, K.; Neda, T.; Kawamura, A.; Monta, K.; Satoh, K.

    1980-01-01

    Recent digital techniques based on changes in electronics and computer technologies have realized a very wide scale of computer application to BWR Power Plant control and instrumentation. Multifarious computers, from micro to mega, are introduced separately. And to get better control and instrumentation system performance, hierarchical computer complex system architecture has been developed. This paper addresses the hierarchical computer complex system architecture which enables more efficient introduction of computer systems to a Nuclear Power Plant. Distributed control and processing systems, which are the components of the hierarchical computer complex, are described in some detail, and the database for the hierarchical computer complex is also discussed. The hierarchical computer complex system has been developed and is now in the detailed design stage for actual power plant application. (auth)

  18. Neutron noise analysis of BWR using time series analysis

    International Nuclear Information System (INIS)

    Fukunishi, Kohyu

    1976-01-01

    The main purpose of this paper is to give more quantitative understanding of noise source in neutron flux and to provide a useful tool for the detection and diagnosis of reactor. The space dependent effects of distributed neutron flux signals at the axial direction of two different strings are investigated by the power contribution ratio among neutron fluxes and the incoherent noise spectra of neutron fluxes derived from autoregressive spectra. The signals are measured on the medium sized commercial BWR of 460 MWe in Japan. From the obtained results, local and global noise sources in neutron flux are discussed. This method is indicated to be a useful tool for detection and diagnosis of anomalous phenomena in BWR. (orig./RW) [de

  19. Recent developments in ultrasonic inspection of BWR pipes in USA

    International Nuclear Information System (INIS)

    Behravesh, M.; Dau, G.J.

    1985-01-01

    Following the 1982 discovery of widespread intergranular stress corrosion cracking (IGSCC) in the recirculation piping system at the Nine Mile Point unit 1 plant, the U.S. Nuclear Regulatory Commission (NRC) made the requirement that the operating BWR plants demonstrate the effectiveness of their ultrasonic testing procedures. Since that time, the detection, identification, and sizing of IGSCC in BWR pipes has received an even greater amount of attention. Initially EPRI through its NDE Center in Charlotte took an active role in providing a mechanism whereby the affected utilities could demonstrate the required capabilities. This effort was soon followed with the development and regular offering of a one week course on IGSCC detection whose final exam was patterned after the requirements of the US NRC Bulletin IE 83-02 on performance capability demonstration

  20. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  1. TVA experience in BWR reload design and licensing

    International Nuclear Information System (INIS)

    Robertson, J.D.

    1986-01-01

    TVA has developed and implemented the capability to perform BWR reload core design and licensing analyses. The advantages accruing from this capability include the tangible cost-savings from performing reload analyses in-house. Also, ''intangible'' benefits such as increased operating flexibility and the ability to accommodate multivendor fuel designs have been demonstrated. The major disadvantage with performing in-house analyses is the cost associated with development and maintenance of the analytical methods and staff expertise

  2. BWR Mark I pressure suppression pool dynamics studies

    International Nuclear Information System (INIS)

    McCauley, E.W.; Martin, R.W.; Lai, W.; Morrison, F.A.; Sutton, S.B.

    1976-11-01

    This report summarizes the initial effort of the Lawrence Livermore Laboratory involvement with the study of BWR Mark I pressure suppression pool dynamics. Analytical activity is described and calculational results are presented for several simplified geometries. Computer code authentication will be provided by a currently active program in benchmark tests. The experiment and some results are presented. A combined analytical and experimental program to evaluate air scaling hypotheses for hydrodynamic forces and pool motion is presented, along with some conclusions regarding model scaling

  3. 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

  4. Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2000-10-12

    Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For criticality safety purposes, current PWR storage and transportation canister designs employ flux traps between assemblies. Credit for fuel burnup will eliminate the need for these flux traps, and thus, significantly increase the PWR assembly capacity (for a fixed canister volume). Increases in assembly capacity of approximately one-third are expected. In contrast, current BWR canister designs do not require flux traps for criticality safety, and thus, are already at their maximum capacity in terms of physical storage. Therefore, benefits associated with burnup credit for BWR storage and transportation casks may be limited to increasing the enrichment capacity and

  5. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  6. Integrated Electronic Microscopy Method to Characterize BWR Crud Deposits

    International Nuclear Information System (INIS)

    Pop, Mike G.; Lockamon, Brian; Howe, James M.; Oleshko, Vladimir P.

    2007-01-01

    The primary objective of this paper is to present the best combination of analytical Electron Microscopy (EM) techniques suited for studying Boiling Water Reactor (BWR) fuel crud deposits in their 'as found' condition, for example un-adulterated portions of the deposits located in crevices at the surface of deposit. The secondary objective of the paper is to suggest a strategy to connect the analytical EM results with the Power Diffraction File (PDF-4) crystal database to ultimately explain crystal growth phenomena in crevices. The samples analyzed as part of this work were collected from steam generators in Nuclear Power Plants. These samples were selected as segregates for BWR deposits due to their similar structures and because they were freely releasable for study in the University labs. The samples were analyzed extensively through various EM techniques at magnifications up to 150,000 X. Subsequent evaluation of the analysis results demonstrated that the selected samples exhibited characteristics that were very close to second-burned fuel crud deposits from operating BWR plants. (authors)

  7. Novel modular natural circulation BWR design and safety evaluation

    International Nuclear Information System (INIS)

    Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

    2015-01-01

    Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

  8. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Radial optimization of a BWR fuel cell using genetic algorithms

    International Nuclear Information System (INIS)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P.

    2006-01-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U 235 and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix the placement of

  10. Improvement in decay ratio calculation in LAPUR5 methodology for BWR instability

    International Nuclear Information System (INIS)

    Li Hsuannien; Yang Tzungshiue; Shih Chunkuan; Wang Jongrong; Lin Haotzu

    2009-01-01

    LAPUR5, based on frequency domain approach, is a computer code that analyzes the core stability and calculates decay ratios (DRs) of boiling water nuclear reactors. In current methodology, one set of parameters (three friction multipliers and one density reactivity coefficient multiplier) is chosen for LAPUR5 input files, LAPURX and LAPURW. The calculation stops and DR for this particular set of parameters is obtained when the convergence criteria (pressure, mass flow rate) are first met. However, there are other sets of parameters which could also meet the same convergence criteria without being identified. In order to cover these ranges of parameters, we developed an improved procedure to calculate DR in LAPUR5. First, we define the ranges and increments of those dominant input parameters in the input files for DR loop search. After LAPUR5 program execution, we can obtain all DRs for every set of parameters which satisfy the converge criteria in one single operation. The part for loop search procedure covers those steps in preparing LAPURX and LAPURW input files. As a demonstration, we looked into the reload design of Kuosheng Unit 2 Cycle 22. We found that the global DR has a maximum at exposure of 9070 MWd/t and the regional DR has a maximum at exposure of 5770 MWd/t. It should be noted that the regional DR turns out to be larger than the global ones for exposures less than 5770 MWd/t. Furthermore, we see that either global or regional DR by the loop search method is greater than the corresponding values from our previous approach. It is concluded that the loop search method can reduce human error and save human labor as compared with the previous version of LAPUR5 methodology. Now the maximum DR can be effectively obtained for a given plant operating conditions and a more precise stability boundary, with less uncertainty, can be plotted on plant power/flow map. (author)

  11. Polygonal instabilities

    Science.gov (United States)

    Labousse, Matthieu

    2015-11-01

    The interaction of a vortex with a free surface is encountered in a series of experiments, the hydraulic jump, the hydraulic bump, the toroidal Leidenfrost experiment. All these experiments share in common an unstable configuration in which azimuthal perturbations give rise to polygonal patterns. We propose a unified theoretical framework to model the emergence of this instability by investigating the stability of a liquid torus with a poloidal motion. As simple as it is, we show that the model retains the necessary ingredients to account for the experimental observations. In this talk, I will first describe the model and compare it to the existing data. However this model is purely inviscid and reaches its limits when being applied to relatively moderate Reynolds flows. So in a second part, I will present a recent experimental and theoretical investigation in which polygonal patterns are now driven by Marangoni flows. To our great surprise, it extends the range of validity of the initial proposed framework, much more than initially expected.

  12. A stability assessment of optimum Fuel Reload Patterns for a BWR

    International Nuclear Information System (INIS)

    Castillo-Durán, Rogelio; Ortiz-Servin, Juan José; Castillo, Alejandro; Montes-Tadeo, José Luis; Perusquía-del-Cueto, Raúl

    2016-01-01

    Highlights: • Fuel reload optimization is made according to steady state and dynamics conditions. • Steady state requirements are thermal limits and cold shutdown margin. • Decay Ratio (DR) is used to measure the core stability. • When DR is not used into the objective function, fuel reloads have high DR values. • Optimized fuel reloads with this methodology have lower DR values. - Abstract: In this paper, a dynamic assessment of optimal BWR Fuel Reload Patterns (FRP) is presented. Using two meta-heuristics techniques: Tabu Search (TS) and Artificial Neural Networks (ANN), the FRPs were designed. In order to perform the stability analysis, a point in the operation map, which is under the instability region defined in technical specifications, was used. The CASMO-4/SIMULATE-3/SIMULATE-3K system was used. The obtained results show that the inclusion of Decay Ratio (DR) into the objective function of TS and ANN optimization techniques, allows finding FRPs that fulfill not only with thermal limits, shutdown margin, energy demand but also with stability requirements. It was possible to achieve FRPs with DR 6% lesser than the corresponding comparison value. The integration of the DR parameter into FRP optimization implied that the needed CPU time to fulfill every requirement was increased by 6.6 times, with respect to the classic steady-state optimization.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  14. Experimental study of the interplay between magmatic rift intrusion and flank instability with application to the 2001 Mount Etna eruption

    KAUST Repository

    Le Corvec, Nicolas

    2014-07-01

    Mount Etna volcano is subject to transient magmatic intrusions and flank movement. The east flank of the edifice, in particular, is moving eastward and is dissected by the Timpe Fault System. The relationship of this eastward motion with intrusions and tectonic fault motion, however, remains poorly constrained. Here we explore this relationship by using analogue experiments that are designed to simulate magmatic rift intrusion, flank movement, and fault activity before, during, and after a magmatic intrusion episode. Using particle image velocimetry allows for a precise temporal and spatial analysis of the development and activity of fault systems. The results show that the occurrence of rift intrusion episodes has a direct effect on fault activity. In such a situation, fault activity may occur or may be hindered, depending on the interplay of fault displacement and flank acceleration in response to dike intrusion. Our results demonstrate that a complex interplay may exist between an active tectonic fault system and magmatically induced flank instability. Episodes of magmatic intrusion change the intensity pattern of horizontal flank displacements and may hinder or activate associated faults. We further compare our results with the GPS data of the Mount Etna 2001 eruption and intrusion. We find that syneruptive displacement rates at the Timpe Fault System have differed from the preeruptive or posteruptive periods, which shows a good agreement of both the experimental and the GPS data. Therefore, understanding the flank instability and flank stability at Mount Etna requires consideration of both tectonic and magmatic forcing. Key Points Analyzing Mount Etna east flank dynamics during the 2001 eruption Good correlation between analogue models and GPS data Understanding the different behavior of faulting before/during/after an eruption © 2014. American Geophysical Union. All Rights Reserved.

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

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1983-01-01

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

  16. Best-practices guidelines for L2PSA development and applications. Volume 2 - Best practices for the Gen II PWR, Gen II BWR L2PSAs. Extension to Gen III reactors

    International Nuclear Information System (INIS)

    Raimond, E.; Durin, T.; Rahni, N.; Meignen, R.; Cranga, M.; Pichereau, F.; Bentaib, A.; Guigueno, Y.; Loeffler, H.; Mildenberger, O.; Lajtha, G.; Santamaria, C.S.; Dienstbier, J.; Rydl, A.; Holmberg, J.E.; Lindholm, I.; Maennistoe, I.; Pauli, E.M.; Dirksen, G.; Grindon, L.; Peers, K.; Hulqvist, G.; Parozzi, F.; Polidoro, F.; Cazzoli, E.; Vitazkova, J.; Burgazzi, L.; Oury, L.; Ngatchou, C.; Siltanen, S.; Niemela, I.; Routamo, T.; Helstroem, P.; Bassi, C.; Brinkman, H.; Seidel, A.; Schubert, B.; Wohlstein, R.; Guentay, S.; Vincon, L.

    2010-01-01

    The objective of this coordinated action was to develop best practice guidelines for the performance of Level 2 PSA methodologies with a view of harmonisation at EU level and to allow meaningful and practical uncertainty evaluations in a Level 2 PSA. Specific relationships with community in charge of nuclear reactor safety (utilities, safety authorities, vendors, and research or services companies) have been established in order to define the current needs in terms of guidelines for level 2 PSA development and applications. An international workshop was organised in Hamburg, with the support of VATTENFALL, in November 2008. The level 2 PSA experts from the ASAMPSA2 project partners have proposed some guidelines for the development and application of L2PSA based on their experience and on information available from international cooperation (EC Severe Accident network of Excellence - SARNET, IAEA standards, OECD-NEA publications and workshop) or open literature. The number of technical issues addressed in the guideline is very large and all are not covered with the same relevancy in the first version of the guideline. This version is submitted for external review in November 2010 by severe accident experts and PSA, especially, from SARNET and OECD-NEA members. The feedback of the external review will be dis cussed during an international open works hop planned in March 2011 and all outcomes will be taken into consideration in the final version of this guideline (June 2011). The guideline includes 3 volumes: - Volume 1 - General considerations on L2PSA. - Volume 2 - Technical recommendations for Gen II and III reactors. - Volume 3 - Specific considerations for future reactor (Gen IV). The recommendations formulated in the guideline should not be considered as 'mandatory' but should help the L2PSA developers to achieve high quality studies with limited time and resources. It may also help the L2PSA reviewers by positioning one specific study in comparison

  17. Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  18. Generic aging management programs for license renewal of BWR reactor coolant systems components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  19. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    1 - Description of program or function: THALES, which stands for 'Thermal Hydraulic Analysis of Loss-of-coolant, Emergency core cooling and Severe core damage', is a computer code system for analyzing progression of core melt accident of light water reactors. The code was developed for Level 2 PSA (probabilistic safety assessment) and applicable to a wide range of postulated accident scenarios. Its outcomes are thermal hydraulic conditions in the reactor coolant system and the containment which are necessary for analyzing fission product release and transport behavior during the accident. The code system consists of following three member codes: (1) THALES-PM for accident progression in the primary and the secondary system of PWRs, (2) THALES-BM for accident progression in the reactor coolant system of BWRs, and (3) THALES-CV for accident progression in the containment of PWRs and BWRs. The THALES-PM and the THALES-BM codes carry out two categories of analysis. The first one is overall thermal-hydraulic analysis in the reactor coolant system. The reactor coolant system is divided into multi-volumes and each volume is further separated into a liquid region and a gas region by a movable mixture level. System pressure, mixture level in each volume, coolant temperature in each region, flow rate between volumes, etc. are calculated. The other one is core heatup and meltdown analysis. The reactor core is radially and axially divided into many nodes. Fuel and cladding temperature, cladding oxidation rate, hydrogen generation rate, core melt fraction, etc. are calculated. The THALES-CV code is for containment response analysis. It divides the containment into multiple compartments, each of which is further separated into a liquid region and a gas region by a movable mixture level. Containment pressure, mixture level in each compartment, coolant temperature in each region, flow rate between compartments, etc. are calculated. The code can treat coolant blowdown from the

  20. Delayed chromosomal instability induced by DNA damage

    International Nuclear Information System (INIS)

    Morgan, W.F.; Marder, B.A.; Day, J.P.

    1994-01-01

    Cellular exposure to DNA damaging agents rapidly results in a dose dependent increase in chromosomal breakage and gross structural chromosomal rearrangements. Over recent years, evidence has been accumulating indicating genomic instability can manifest multiple generations after cellular exposure to physical and chemical DNA damaging agents. Genomic instability manifests in the progeny of surviving cells, and has been implicated in mutation, gene application, cellular transformation, and cell killing. To investigate chromosome instability following DNA damage, we have used fluorescence in situ hybridization to detect chromosomal rearrangements in a human/hamster somatic hybrid cell line following exposure to ionizing radiation. Delayed chromosomal instability was detected when multiple populations of uniquely arranged metaphases were observed in clonal isolates raised from single cells surviving X-irradiation many generations after exposure. At higher radiation doses, chromosomal instability was observed in a relatively high frequency of surviving clones and, in general, those clones showed delayed chromosome instability also showed reduced survival as measured by colony forming ability

  1. Determination of the equivalent intergranular void ratio - Application to the instability and the critical state of silty sand

    Directory of Open Access Journals (Sweden)

    Nguyen Trung-Kien

    2017-01-01

    Full Text Available This paper presents an experimental study of mechanical response of natural Camargue silty sand. The analysis of test results used the equivalent intergranular void ratio instead of the global void ratio. The calculation of equivalent intergranular void ratio requires the determination of parameter b which represents, physically, the fraction of active fines participating on the chain forces network, hence the strength of the soil. A new formula for determining the parameter b by using an approach based on the coordination number distribution and probability calculation is proposed. The validation of the developed relationship was done through back-analysis of published datasets in literature on the effect of fines content on silty sand behavior. It is shown that the equivalent intergranular void ratio calculated with the b value obtained by the new formula is able to provide strong correlation to not only the critical state of but also the onset of instability of various silty sands, in different terms as peak deviator stress, peak stress ratio or cyclic resistance. Therefore, it is suggested that the use of the equivalent void ratio concept and the new b calculating formula is highly desirable in predicting of the silty sand behavior.

  2. Prevention of organic iodide formation in BWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Karjunen, T. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Laitinen, T.; Piippo, J.; Sirkiae, P. [VTT Manufacturing Technology (Finland)

    1996-12-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR`s as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs.

  3. 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)

  4. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

    International Nuclear Information System (INIS)

    Gonzalez C, J.; Martin del Campo M, C.

    2003-01-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  5. The mechanical structure of the SVEA BWR fuel

    International Nuclear Information System (INIS)

    Nylund, O.; Johansson, A.; Junkrans, S.

    1985-01-01

    The SVEA BWR fuel assembly design is characterized by a double-wall cruciform internal structure forming an internal water gap and dividing the assembly into 4 subbundles. The effect is a favourable distribution of fuel and moderator, a minimum amount of structural material in active core, a combination of structural stability and flexibility for minimum control rod friction in reduced gaps and a reduced creep deformation of the fuel assembly. The results of a laboratory test program confirm the much lower friction force obtained with the SVEA fuel assemblies while withdrawing and inserting the control rod. (RF)

  6. 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)

  7. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  8. Improving BWR fuel critical power without increasing bundle pressure drop

    International Nuclear Information System (INIS)

    Matzner, B.; Shiraishi, L.M.; Danielson, D.W.; Congdon, S.P.

    2004-01-01

    It has been almost axiomatic that BWR fuel bundle critical power performance could not be improved without an accompanying increase in bundle pressure drop. It appeared that in order to increase the bundle dryout resistance it was necessary to perturb the bundle coolant flow paths in some fashion. This resulted in an unacceptable bundle pressure drop increase. However, by adding part length rods to decrease bundle pressure drop and by inserting an extra spacer with rearranged spacer pitch and flow trippers on the channel wall at the top of the bundle to increase critical power it was possible to achieve the goal of increased bundle critical power without pressure drop increase. (author)

  9. In-situ testing of BWR closure head studs

    International Nuclear Information System (INIS)

    deRaad, J.A.; Wolters, J.T.

    1988-01-01

    Mechanized ultrasonic inspection of closure head studs often is on the critical path. In German BWR's, a floodcompensator is used which allows human access to the studs despite the water is up to a much higher level. For stud inspection this provides a potential solution to get out of the critical path. However, the space restrictions around the studs due to the geometry of the floodcompensator did not allow the use of the existing manipulators. This paper describes the design of a dedicated compact manipulator of a construction which copes with the restricted space available around the studs

  10. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  11. Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State

    2016-11-30

    The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup

  12. Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 nuclear island design (Docket No. 50-447). Supplement No. 4

    International Nuclear Information System (INIS)

    1985-07-01

    Supplement 4 to the Safety Evaluation Report (SER) for the application filed by General Electric Company for the final design approval for the GE BWR/6 nuclear island design (GESSAR II) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report supplements the GESSAR II SER (NUREG-0979) issued in April 1983 summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design; Supplement 1, issued in July 1983; Supplement 2, issued in November 1984; and Supplement 3, issued in January 1985. Subject to favorable resolution of the items discussed in this supplement, the staff concludes that the GESSAR II design satisfactorily addresses the severe-accident concerns described in the Commission's Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants

  13. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    Schienbein, Marcel; Zeh, Peter; Hurtado, Antonio; Rosskamp, Matthias; Mailand, Irene; Bolz, Michael

    2012-09-01

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  14. Failure frequencies and probabilities applicable to BWR and PWR piping

    International Nuclear Information System (INIS)

    Bush, S.H.; Chockie, A.D.

    1996-03-01

    This report deals with failure probabilities and failure frequencies of nuclear plant piping and the failure frequencies of flanges and bellows. Piping failure probabilities are derived from Piping Reliability Analysis Including Seismic Events (PRAISE) computer code calculations based on fatigue and intergranular stress corrosion as failure mechanisms. Values for both failure probabilities and failure frequencies are cited from several sources to yield a better evaluation of the spread in mean and median values as well as the widths of the uncertainty bands. A general conclusion is that the numbers from WASH-1400 often used in PRAs are unduly conservative. Failure frequencies for both leaks and large breaks tend to be higher than would be calculated using the failure probabilities, primarily because the frequencies are based on a relatively small number of operating years. Also, failure probabilities are substantially lower because of the probability distributions used in PRAISE calculations. A general conclusion is that large LOCA probability values calculated using PRAISE will be quite small, on the order of less than 1E-8 per year (<1E-8/year). The values in this report should be recognized as having inherent limitations and should be considered as estimates and not absolute values. 24 refs 24 refs

  15. Failure frequencies and probabilities applicable to BWR and PWR piping

    Energy Technology Data Exchange (ETDEWEB)

    Bush, S.H. [Review and Synthesis Associates, Richland, WA (United States); Chockie, A.D. [Chockie Group International Inc., Seattle, WA (United States)

    1996-03-01

    This report deals with failure probabilities and failure frequencies of nuclear plant piping and the failure frequencies of flanges and bellows. Piping failure probabilities are derived from Piping Reliability Analysis Including Seismic Events (PRAISE) computer code calculations based on fatigue and intergranular stress corrosion as failure mechanisms. Values for both failure probabilities and failure frequencies are cited from several sources to yield a better evaluation of the spread in mean and median values as well as the widths of the uncertainty bands. A general conclusion is that the numbers from WASH-1400 often used in PRAs are unduly conservative. Failure frequencies for both leaks and large breaks tend to be higher than would be calculated using the failure probabilities, primarily because the frequencies are based on a relatively small number of operating years. Also, failure probabilities are substantially lower because of the probability distributions used in PRAISE calculations. A general conclusion is that large LOCA probability values calculated using PRAISE will be quite small, on the order of less than 1E-8 per year (<1E-8/year). The values in this report should be recognized as having inherent limitations and should be considered as estimates and not absolute values. 24 refs 24 refs.

  16. BWR noise spectra and application of noise analysis to FBR

    International Nuclear Information System (INIS)

    Nomura, T.

    1975-01-01

    Work related to noise analysis, in Tokyo Shibaura Electric Co. Ltd. (Toshiba) and Nippon Atomic Industry Group Co. Ltd. (NAIG) for the past several years is reviewed. After considering the Japan-United States Seminar on Reactor Noise Analysis in 1968, other subjects discussed were boiling water reactor noise analysis and work in relation to FBR. Parts of these are related to each other. For example, boiling detection and temperature fluctuations are problems pertinent to both fields. As the main problems in zero-power-reactor noise are now basically understood, only a brief description of the experiments involving the advanced two detector method is made. Focus is rather placed on the area of power plant noise. (author)

  17. Automatic determination of BWR fuel loading patterns based on K.E. technique with core physics simulation

    International Nuclear Information System (INIS)

    Ikehara, T.; Tsuiki, M.; Takeshita, T.

    1990-01-01

    On the basis oof a computerized search method, a prototype for a fuel loading pattern expert system has been developed to support designers in core design for BWRs. The method was implemented by coupling rules and core physics simulators into an inference engine to establish an automated generate-and-test cycle. A search control mechanism, which prunes paths to be searched and selects appropriate rules through the interaction with the user, was also introduced to accomplish an effective search. The constraints in BWR core design are: (1) cycle length more than L, (2) core shutdown margin more than S, and (3) thermal margin more than T. Here L, S, and T are the specified minimum values. In this system, individual rules contain the manipulation to improve the core shutdown margin explicitly. Other items were taken into account only implicitly. Several applications to the test cases were carried out. It was found that the results were comparable with those obtained by human expert engineers. Broad applicability of the present method in the BWR core design domain was proved

  18. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  19. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  20. Safety evaluation of BWR off-gas treatment systems

    International Nuclear Information System (INIS)

    Schultz, R.J.; Schmitt, R.C.

    1975-01-01

    Some of the results of a safety evaluation performed on current generic types of BWR off-gas treatment systems including cooled and ambient temperature adsorber beds and cryogenics are presented. The evaluation covered the four generic types of off-gas systems and the systems of five major vendors. This study was part of original work performed under AEC contract for the Directorate of Regulatory Standards. The analysis techniques employed for the safety evaluation of these systems include: Fault Tree Analysis; FMECA (Failure Mode Effects and Criticality Analysis); general system comparisons, contaminant, system control, and design adequacy evaluations; and resultant Off-Site Dose Calculations. The salient areas presented are some of the potential problem areas, the approach that industry has taken to mitigate or design against potential upset conditions, and areas where possible deficiencies still exist. Potential problem areas discussed include hydrogen detonation, hydrogen release to equipment areas, operator/automatic control interface, and needed engineering evaluation to insure safe system operation. Of the systems reviewed, most were in the category of advanced or improved over that commonly in use today, and a conclusion from the study was that these systems offer excellent potential for noble gas control for BWR power plants where more stringent controls may be specified -- now or in the future. (U.S.)

  1. Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

    International Nuclear Information System (INIS)

    Cole, Steven E.; Garner, Norman L.; Lippert, Hans-Joachim; Graebert, Rüdiger; Mollard, Pierre; Hahn, Gregory C.

    2014-01-01

    Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

  2. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs

  3. An optimized BWR fuel lattice for improved fuel utilization

    International Nuclear Information System (INIS)

    Bernander, O.; Helmersson, S.; Schoen, C.G.

    1984-01-01

    Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

  4. Logical model for the control of a BWR turbine

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  5. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.; Farmer, W.S.

    1992-01-01

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  6. Sophistication of operator training using BWR plant simulator

    International Nuclear Information System (INIS)

    Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

    1986-01-01

    In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

  7. Radiation buildup and control in BWR recirculation piping

    International Nuclear Information System (INIS)

    Meyer, W.; Wood, R.M.; Rao, T.V.; Vook, R.W.

    1987-01-01

    Boiling water nuclear reactors (BWRs) employ stainless steel (Types 304 or 316 NG) pipes in which high-purity water at temperatures of ∼ 275 0 C are circulated. Various components of the system, such as valves and bearings, often contain hard facing metal alloys such as Stellite-6. These components, along with the stainless steel tubing and feedwater, serve as sources of 59 Co. This cobalt, along with other soluble and insoluble impurities, is carried along with the circulating water to the reactor core where it is converted to radioactive 60 Co. After reentering the circulating water, the 60 Co can be incorporated into a complex corrosion layer in the form of CoCr 2 O 4 and/or CoFe 2 O 4 . The presence of even small amounts of 60 Co on the walls of BWR cooling systems is the dominant contributor to inplant radiation levels. Thus BWR owners and their agents are expending significant time and resources in efforts to reduce both the rate and amount of 60 Co buildup. The object of this research is twofold: (a) to form a thin diffusion barrier against the outward migration of cobalt from a cobalt-containing surface and (b) to prevent the growth of a 60 Co-containing corrosion film. The latter goal was the more important since most of the radioactive cobalt will originate from sources other than the stainless steel piping itself

  8. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  9. System control model of a turbine for a BWR

    International Nuclear Information System (INIS)

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

    2009-10-01

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

  10. Utility experience with the BWR power shape monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Emrich, W.J. Jr.

    1986-01-01

    The Oyster Creek Nuclear Generating Station has used the Electric Power Research Institute (EPRI) boiling water reactor (BWR) power shape monitoring system (PSMS) for approximately one cycle of power operation. During this time, the PSMS was used for both on-line core monitoring and off-line core simulation. The ability to obtain core power distributions in near real time and the ability to simulate anticipated core maneuvers has enabled site personnel to avoid difficult operational situations and to more easily detect problems when they occur. The PSMS is an integrated set of computer codes used to monitor and predict the nuclear and thermal-hydraulic behavior of a BWR code. The PSMS uses a three-dimensional, one-energy-group nodal model similar to the General Electric FLARE code to estimate the core power distribution. The PSMS contains four major modules: (1) data acquisition and storage module: collects plant data; (2) duty cycle approximation module; performs three-dimensional nuclear and thermal-hydraulic calculations; (3) model calibration module: performs all statistical evaluations; and (4) data display module: responsible for displaying PSMS data. The PSMS also has the ability to perform useful predictive calculations. These include: power, flow, or control rod searches; control rod pattern optimizations; and fast approximate predictions.

  11. Fringe instability in constrained soft elastic layers.

    Science.gov (United States)

    Lin, Shaoting; Cohen, Tal; Zhang, Teng; Yuk, Hyunwoo; Abeyaratne, Rohan; Zhao, Xuanhe

    2016-11-04

    Soft elastic layers with top and bottom surfaces adhered to rigid bodies are abundant in biological organisms and engineering applications. As the rigid bodies are pulled apart, the stressed layer can exhibit various modes of mechanical instabilities. In cases where the layer's thickness is much smaller than its length and width, the dominant modes that have been studied are the cavitation, interfacial and fingering instabilities. Here we report a new mode of instability which emerges if the thickness of the constrained elastic layer is comparable to or smaller than its width. In this case, the middle portion along the layer's thickness elongates nearly uniformly while the constrained fringe portions of the layer deform nonuniformly. When the applied stretch reaches a critical value, the exposed free surfaces of the fringe portions begin to undulate periodically without debonding from the rigid bodies, giving the fringe instability. We use experiments, theory and numerical simulations to quantitatively explain the fringe instability and derive scaling laws for its critical stress, critical strain and wavelength. We show that in a force controlled setting the elastic fingering instability is associated with a snap-through buckling that does not exist for the fringe instability. The discovery of the fringe instability will not only advance the understanding of mechanical instabilities in soft materials but also have implications for biological and engineered adhesives and joints.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stellwag, Bernhard; Laendner, Alexander; Weiss, Steffen [AREVA NP GmbH, Erlangen (Germany); Huettner, Frank [Vattenfall Europe Nuclear Energy GmbH (Germany)

    2011-03-15

    Nineteen boiling water reactor (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 the 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. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Stellwag, B.; Laendner, A.; Weiss, S. [AREVA NP GmbH, Erlangen (Germany); Huettner, F. [Vattenfall Europe Nuclear Energy GmbH, Hamburg (Germany)

    2010-07-01

    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)

  15. Optimization of fuel cells for BWR based in Tabu modified search; Optimizacion de celdas de combustible para BWR basada en busqueda Tabu modificada

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Palomera P, M.A. [Facultad de Ingenieria, UNAM, Posgrado en Ingenieria en Computacion, Circuito exterior s/n, Ciudad Universitaria, Mexico, D.F. (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2004-07-01

    The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  17. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

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

  19. Treatment of the subject of tearing instability

    International Nuclear Information System (INIS)

    Paris, P.C.

    1977-07-01

    A simple approach is taken to the mechanics of potential instability associated with the steady tearing portion of J-Integral R-curves. The analysis is developed from simple examples of structural component (or test specimen) configurations with cracks, examining their instability possibilities individually, in order to draw more general conclusions about elastic-plastic cracking instability as contrasted to linear-elastic behavior. Finally, an attempt is made to model a more local cleavage-like instability for material in the fracture process zone just ahead of a crack tip. Results are then presented of a testing program which clearly demonstrates the appropriateness of the tearing instability analysis and which illustrates its broad potential for future application, as well as presenting guidelines for its further development. The material selected for analysis was Ni-Cr-Mo-V rotor steel

  20. The DNA-instability test as a specific marker of malignancy and its application to detect cancer clones in borderline malignancy

    Directory of Open Access Journals (Sweden)

    M Fukuda

    2009-06-01

    Full Text Available Recent progress in cytogenetic and biochemical mutator assay technologies has enabled us to detect single gene alterations and gross chromosomal rearrangements, and it became clear that all cancer cells are genetically unstable. In order to detect the genome-wide instability of cancer cells, a new simple method, the DNA-instability test, was developed. The methods to detect genomic instability so far reported have only demonstrated the presence of qualitative and quantitative alterations in certain specific genomic loci. In contrast to these commonly used methods to reveal the genomic instability at certain specific DNA regions, the newly introduced DNA-instability test revealed the presence of physical DNA-instability in the entire DNA molecule of a cancer cell nucleus as revealed by increased liability to denature upon HCl hydrolysis or formamide exposure. When this test was applied to borderline malignancies, cancer clones were detected in all cases at an early-stage of cancer progression. We proposed a new concept of “procancer” clones to define those cancer clones with “functional atypia” showing positivities for various cancer markers, as well as DNA-instability testing, but showing no remarkable ordinary “morphological atypia” which is commonly used as the basis of histopathological diagnosis of malignancy.

  1. A low-cost, label-free DNA detection method in lab-on-chip format based on electrohydrodynamic instabilities, with application to long-range PCR.

    Science.gov (United States)

    Diakité, Mohamed Lemine Youba; Champ, Jerôme; Descroix, Stephanie; Malaquin, Laurent; Amblard, François; Viovy, Jean-Louis

    2012-11-21

    In order to evolve from a "chip in the lab" to a "lab on a chip" paradigm, there is still a strong demand for low-cost, portable detection technologies, notably for analytes at low concentrations. Here we report a new label-free DNA detection method with direct electronic read, and apply it to long-range PCR. This method uses a nonlinear electrohydrodynamic phenomenon: when subjected to high electric fields (typically above 100 V cm(-1)), suspensions of large polyelectrolytes, such as long DNA molecules, create "giant" dynamic concentration fluctuations. These fluctuations are associated with large conductivity inhomogeneities, and we use here a contact-mode local conductivity detector to detect these fluctuations. In order to decouple the detection electronics from the high voltage excitation one, an original "doubly symmetric" floating mode battery-operated detection scheme was developed. A wavelet analysis was then applied, to unravel from the chaotic character of the electohydrodynamic instabilities a scalar signal robustly reflecting the amplification of DNA. As a first proof of concept, we measured the products of the off-chip amplification of 10 kbp DNA from lambda phage DNA, achieving a sensitivity better than 100 fg DNA in the original 50 μl sample. This corresponds to the amplification products of less than 100 initial copies of target DNA. The companion enabling technologies developed to implement this new concept, i.e. the doubly symmetric contact conductivity detection and wavelet analysis, may also find various other applications in lab-on-chips.

  2. Obtention control bars patterns for a BWR using Tabo search

    International Nuclear Information System (INIS)

    Castillo, A.; Ortiz, J.J.; Alonso, G.; Morales, L.B.; Valle, E. del

    2004-01-01

    The obtained results when implementing the technique of tabu search, for to optimize patterns of control bars in a BWR type reactor, using the CM-PRESTO code are presented. The patterns of control bars were obtained for the designs of fuel reloads obtained in a previous work, using the same technique. The obtained results correspond to a cycle of 18 months using 112 fresh fuels enriched at the 3.53 of U-235. The used technique of tabu search, prohibits recently visited movements, in the position that correspond to the axial positions of the control bars, additionally the tiempo t abu matrix is used for to manage a size of variable tabu list and the objective function is punished with the frequency of the forbidden movements. The obtained patterns of control bars improve the longitude of the cycle with regard to the reference values and they complete the restrictions of safety. (Author)

  3. TRAB - A transient analysis program for BWR. Part 2

    International Nuclear Information System (INIS)

    Raety, H.; Rajamaeki, M.

    1991-05-01

    TRAB is a transient analysis code for BWRs developed at the Technical Research Centre of Finland. It models the phenomena in the interior of the BWR pressure vessel and in related subsystems. The core model of TRAB can be used separately for LWR modelling. For PWR modelling the core model of TRAB is connected to circuit model SMABRE to form the SMATRA code. This report is a user's manual and documents the structure, contents and preparation of input for TRAB. The structure of TRAB input is very flexible, featuring input groups and subgroups identified with keywords and given in any order as well as data items in free format, freely mixed with explanatory texts. Users interface of the code can be used for modelling within input: through normal input it is possible to create new submodels. These may be functional or tabulated dependencies of the code variables, different types of delays, or ordinary linear differential equations

  4. Seismic response analysis of BWR buildings with embedded foundation

    International Nuclear Information System (INIS)

    Fukuzawa, R.; Chiba, O.; Tohdo, M.

    1985-01-01

    An investigation on the effect of various types of foundations embedded into soil deposit to the responses of BWR buildings is made. The frequency characteristics of massless rigid foundations such as impedance functions and input motion are first obtained, depending on the embedment depth, contact condition of side wall to soil and the thickness of surface layer in a two-layered soil. From the results it is found that the embedment depth and the thickness of surface layer have a distinct effect to the responses contributed by the impedance functions and the response values characterized by the input motions are obviously influenced by the embedment depth and the contact length of side wall to soil deposit. (orig.)

  5. BWR Mark I pressure suppression study: bench mark experiments

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1977-01-01

    Computer simulations representative of the wetwell of Mark I BWR's have predicted pressures and related phenomena. However, calculational predictions for purposes of engineering decision will be possible only if the code can be verified, i.e., shown to compute in accord with measured values. Described in the report is a set of single downcomer spherical flask bench mark experiments designed to produce quantitative data to validate various air-water dynamic computations; the experiments were performed since relevant bench mark data were not available from outside sources. Secondary purposes of the study were to provide a test bed for the instrumentation and post-experiment data processing techniques to be used in the Laboratory's reactor safety research program and to provide additional masurements for the air-water scaling study

  6. Study of transient turbine shot without bypass in a BWR

    International Nuclear Information System (INIS)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

    2015-09-01

    The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

  7. Preliminary study on seismic proving test of BWR core internals

    International Nuclear Information System (INIS)

    Watabe, M.; Shibata, H.; Sato, H.; Kawakami, S.; Ohno, T.; Ichiki, T.; Sasaki, Y.

    1989-01-01

    A series of seismic proving tests for PWR and BWR facilities have been programmed and conducted. The seismic design of nuclear plants is based on criteria far more stringent than applied to non-nuclear installations. Seismic design involves the complex, advanced and sophisticated calculations. The validity of the techniques employed for these calculations is not susceptible to easy comprehension and acceptance by the population at large. It has become important to demonstrate the seismic strength and functional reliability of currently designed and existing nuclear power facilities. This paper deals with an experimental research project to demonstrate the seismic reliability proving tests for nuclear power plant facilities. The seismic proving test described in this report was conducted from 1986 to 1988 for the boiling water reactor core internals

  8. BWR plant analyzer development at BNL (Brookhaven National Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

    An engineering plant analyzer has been developed at BNL for realistically and accurately simulating transients and severe abnormal events in BWR power plants. Simulations are being carried out routinely with high fidelity, high simulation speed, at low cost and with unsurpassed user convenience. The BNL Plant Analyzer is the only operating facility which (a) simulates more than two orders-of-magnitude faster than the CDC-7600 mainframe computer, (b) is accessible and fully operational in on-line interactive mode, remotely from anywhere in the US, from Europe or the Far East (Korea), via widely available IBM-PC compatible personal computers, standard modems and telephone lines, (c) simulates both slow and rapid transients seven times faster than real-time in direct access, and four times faster in remote access modes, (d) achieves high simulation speed without compromising fidelity, and (e) is available to remote access users at the low cost of $160 per hour.

  9. Blowdown heat transfer and transient boiling transition in BWR's

    International Nuclear Information System (INIS)

    Sozzi, G.L.; Burnette, G.W.

    1977-01-01

    Experimental results from the NRC/EPRI/GE BWR Blowdown Heat Transfer Program are evaluated in terms of bundle local heat transfer performance and in terms of cross-sectional average bundle thermal-hydraulic fluid conditions. The bundle heat transfer performance was generally found to be nucleate boiling below the two-phase mixture level interface with highly dispersed film boiling or steam cooling heat transfer above the interface. Comparisons are presented for predictions of boiling transition (BT) and post BT heat transfer performance during the blowdown phase of the LOCA experiments. These predictions utilize a drift flux void fraction model. The comparisons show very good agreement of both the onset of BT and the post BT heat transfer. 12 references

  10. 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)

  11. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Uruma, Y.; Osato, T.; Yamazaki, K.

    2002-01-01

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  12. Joint Instability and Osteoarthritis

    Directory of Open Access Journals (Sweden)

    Darryl Blalock

    2015-01-01

    Full Text Available Joint instability creates a clinical and economic burden in the health care system. Injuries and disorders that directly damage the joint structure or lead to joint instability are highly associated with osteoarthritis (OA. Thus, understanding the physiology of joint stability and the mechanisms of joint instability-induced OA is of clinical significance. The first section of this review discusses the structure and function of major joint tissues, including periarticular muscles, which play a significant role in joint stability. Because the knee, ankle, and shoulder joints demonstrate a high incidence of ligament injury and joint instability, the second section summarizes the mechanisms of ligament injury-associated joint instability of these joints. The final section highlights the recent advances in the understanding of the mechanical and biological mechanisms of joint instability-induced OA. These advances may lead to new opportunities for clinical intervention in the prevention and early treatment of OA.

  13. Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

    developed work only concentrates in the reactor core, but taking advantage of the modularity that PVM offers, it is possible to add component such as separators and steam dryers, lines of steam and feed water to obtain a model of a complete closed circuit. The applications concentrate mainly in the training of personnel in the phenomenology of the BWR, and as an investigation tool in the study of the dynamics of BWR reactors. The oscillations out of phase study presents challenges at the moment as are the explanation of the variation of the neutral line with time, non azimuthal but axial oscillations out of phase, etc. So far a first model oriented in this direction is at hand. [Spanish] Este trabajo consiste en la integracion de tres modelos desarrollados previamente los cuales se encuentran ampliamente descritos en la literatura: modelo del canal termohidraulico, modelo de la neutronica modal y el modelo de los lazos de recirculacion. La herramienta utilizada para este acoplamiento de modelos es el sistema PVM, Parallel Virtual Machine, que permitio paralelizar el modelo mediante el concepto de computacion distribuida. La finalidad de hacer este acoplamiento de modelos es la de obtener una herramienta mas completa que represente mejor la configuracion real y la fenomenologia del nucleo de un reactor BWR, obteniendo asi mejores resultados. Ademas mantener la flexibilidad de mejorar el modelo resultante en cualquier momento, ya que los modelos muy complejos o sofisticados resultan dificiles de mejorar siendo imposible modificar las ecuaciones que utilizan y pueden incluir variables que no son de importancia primaria en el problema tratado o que enmascaren relaciones entre variables debido al exceso de resultados. Tambien el mantener la flexibilidad de agregar modelos de componentes o sistemas del reactor BWR, todo esto dependiendo de las necesidades del modelado. Se eligio a la planta sueca Ringhals para caracterizar el modelo acoplado resultante por contar con un Benchmark

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

    International Nuclear Information System (INIS)

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

    2002-12-01

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

  15. Hydrodynamic instabilities in inertial fusion

    International Nuclear Information System (INIS)

    Hoffman, N.M.

    1994-01-01

    This report discusses topics on hydrodynamics instabilities in inertial confinement: linear analysis of Rayleigh-Taylor instability; ablation-surface instability; bubble rise in late-stage Rayleigh-Taylor instability; and saturation and multimode interactions in intermediate-stage Rayleigh-Taylor instability

  16. Neutrino beam plasma instability

    Indian Academy of Sciences (India)

    We derive relativistic fluid set of equations for neutrinos and electrons from relativistic Vlasov equations with Fermi weak interaction force. Using these fluid equations, we obtain a dispersion relation describing neutrino beam plasma instability, which is little different from normal dispersion relation of streaming instability.

  17. Analysis of the behavior of irradiated BWR fuel rod in storage dry conditions; Analisis del comportamiento de una barra combustible irradiada BWR en condiciones de almacenamiento en seco

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, A.; Montes, D.; Ruiz-Hervias, J.; Munoz-Reja, C.

    2014-07-01

    In order to complete previous studies of creep on PWR sheath material, developed a joint experimental program by CSN, ENRESA and ENUSA about BWR (Zircaloy-2) sheath material. This program consisted in creep tests and then on the material under creep, compression testing diametral obtaining the permissible displacement of the sheath to break. (Author)

  18. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS.

  19. Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS.

  20. Microstructure and corrosion studies for optimized PWR and BWR Zircaloy cladding

    International Nuclear Information System (INIS)

    Garzarolli, F.; Steinberg, E.; Weidinger, H.G.

    1989-01-01

    The increased discharge burnups envisaged today require an optimization of the Zircaloy cladding properties with regard to corrosion with different strategies for optimizing the corrosion of boiling water reactor (BWR) and pressure water reactor (PWR) applications. The authors report that a broad laboratory study showed that the influence of temperature and environmental details upon autoclave test results is rather complex. Zircaloy with fine (∼ 0.05 μm), uniformly distributed intermetallic precipitates reveals nodular corrosion in steam at 420 degrees C or below and excessive uniform corrosion in water at 350 degrees C, but the best nodular behavior in steam at 500 degrees C. However, samples with larger intermetallic precipitates (> 0.2 μm) exhibit a strong tendency to nodular corrosion in 500 degrees C steam, but good behavior in steam at 420 degrees C or below and in water at 350 degrees C. The tendency toward nodular corrosion in 400 to 500 degrees C steam can be suppressed by additions of oxygen; the ranking in uniform corrosion in 350 degrees C water can be altered by additions of hydrogen. The latter was observed on samples exposed up to 1000 days in degassed water and afterwards 1000 days in water with a hydrogen overpressure

  1. Optimization of fuel cells for BWR based in Tabu modified search

    International Nuclear Information System (INIS)

    Martin del Campo M, C.; Francois L, J.L.; Palomera P, M.A.

    2004-01-01

    The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  3. Increase of Steam Moisture in the BWR-Facility KKP 1

    International Nuclear Information System (INIS)

    Noack, Volker

    2002-01-01

    Main steam moisture in a BWR facility is determined by steam quality at core outlet and efficiency of steam separators and steam dryers. Transport of water with steam is accompanied by transport of radionuclides out of RPV resulting in enhanced radiation level in the main steam system. A remarkable increase of main steam moisture started at KKP 1 in 1997. In the following years increase of steam outlet moisture started at lower and lower core mass flow rates. Dose rate in main steam system increased simultaneously. Core mass flow rate and thus thermal power had to be reduced during stretch out operation to keep the main steam moisture below the specified boundary of 0.2 %. This boundary also guarantees, that radiological exposure remains far below approved values. The increase of main steam moisture corresponds with the application of low leakage core loading. Low leakage core loading results in enhanced steam generation in the center and in reduced steam generation in the outer zones of the core. It can be shown, that the uneven steam generation in the core became stronger over the years. Therefore, steam quality at inlet of the outer steam separators was getting lower. This resulted in higher carry over of water in this steam separators and steam dryers, thus explaining the increasing main steam moisture. KKP 1 started in 2000 with spectral shift operation. As one should expect, this resulted in reduced steam moisture. It remains the question of steam moisture in case of stretch out operation. Countermeasures are briefly discussed. (authors)

  4. An Investigation of Pool Stratification Effects on BWR Wetwell External Venting Using GOTHIC

    International Nuclear Information System (INIS)

    George, T.L.; Harrell, J.R.; Marshall, M.D.; Ozdemir, O.E.; Massin, S.; Rubio, G.

    2017-01-01

    Strategies for coping with severe accidents and long term loss of AC power at BWR nuclear plants may include the use of wetwell and drywell vents to the atmosphere. In some forensic analyses for the Fukushima events, stratification of the wetwell has been postulated as a contributing factor to the higher than expected pressurization rate of the containment. The GOTHICh code has been used to investigate the containment response during the Fukushima events, including the possibility of suppression pool stratification. Thermal stratification, if it develops, would diminish the heat absorbing capacity of the pool, leading to earlier and more frequent vent valve cycling and earlier steam bypass through the saturated pool. This paper describes the use of the GOTHIC computer code to investigate the containment response during a long term loss of AC power with periodic wetwell venting, including the possibility of suppression pool stratification. The applicability of GOTHIC to pool stratification is discussed and modeling strategies are described for dealing with the long transients that must be considered for vent performance. GOTHIC results are presented for one selected loss of AC power scenario with and without stratification effects included. (author)

  5. Westinghouse PWR and BWR reactor vessel segmentation experience in using mechanical cutting process

    International Nuclear Information System (INIS)

    Segerud, Per; Fallstroem, Stefan; Boucau, Joseph; Kreitman, Paul J.

    2011-01-01

    Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR's) and Boiling Water Reactors (BWR's) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date. (author)

  6. A Review of LOX/Kerosene Combustion Instability in American and Russian Combustion Devices in Application to Next-Generation Launch Technology

    Science.gov (United States)

    Rocker, Marvin; Nesman, Tomas E.; Hulka, James R.; Dougherty, N. Sam

    2003-01-01

    The Next-Generation Launch Technology (NGLT) project was introduced with its objectives. To meet the objectives, NASA has directed aerospace industry to perform advances and risk reduction of relevant technologies, including propulsion. Originally, the propulsion industry focused on producing both LOWLH2 and LOWkerosene flight engine technology demonstrators. These flight engine technology demonstrators were briefly reviewed. NASA recently redirected this focus to Lowkerosene only. Discussion of LOWkerosene combustion devices was and is prefaced by grave concerns about combustion instability. These concerns have prompted a review of LOWkerosene combustion instability in American and Russian combustion devices. In the review of the Russian propulsion industry's experience in eliminating LOWkerosene combustion instabilities, the history of principal Russian rocket scientists and their role in the development of LOXkerosene combustion devices is presented. The innovative methods implemented by the Russians of eliminations combustion instabilities in LOXkerosene combustion devices were reviewed. The successful elimination of these combustion instabilities has resulted in two generations of Russian-produced, high-performance LOWkerosene combustion devices.

  7. Instability of refrigeration system - A review

    International Nuclear Information System (INIS)

    Liang Nan; Shao Shuangquan; Xu Hongbo; Tian Changqing

    2010-01-01

    It is essential to ensure the stability for the normal operation of refrigeration systems. This paper reviews the researches on the theory and solutions of the instability of refrigeration systems. The instability of refrigeration systems includes two aspects: the two-phase flow instability in refrigeration system, the instability on refrigeration system control characteristics. As an inherent characteristic of two-phase evaporating flow, several separate explanations for the formation of oscillation of mixture-vapor transition point in the evaporation process by different scholars had been given but there is no general explanation till now. The investigation of instability on refrigeration system control characteristics focused on both static and dynamic researches. The minimum stable signal line theory, as a very important finding for the static instability of the evaporator and thermal expansion valve control loop, presented the different result to other researches. Dynamic researches on simulation and frequency-domain analysis provided various means for forecast and validation with considerable precision while their application range was still confined. With the development of variable capacity compressor and electronic expansion valve, further researches should be carried out to analyze the instability of the variable capacity refrigeration system with considering the influence of parameter coupling and control algorithm.

  8. Study of the possibility of localising a channel instability in Forsmark-1

    International Nuclear Information System (INIS)

    Karlsson, J.K-H.; Pazsit, I.

    1998-01-01

    A special type of instability occurred in the Swedish BWR Forsmark 1 in 1996. In contrast to the better known global or regional instabilities, the decay ratio appeared to be very high in one half of the core and quite low in the other half. A more detailed analysis showed that the most likely reason for the observed behaviour is a local perturbation of thermohydraulic character induced by the incorrect positioning of a fuel assembly. In such a case it is of importance to determine the position of the unseated assembly already during operation so that it can be easily found during reloading. The subject of this paper is to report on development and application of a method by which the position of such a local perturbation can be determined. The method can be separated into two parts that support and complement each other. First a visualisation technique was elaborated which displays the space-time behaviour of the neutron flux oscillations in the core. This visualisation expedites a very good qualitative comprehension of the situation and can be useful for the operators. It also gives a basis for the application of the localisation algorithm. Second, a quantitative localisation method, based on reactor physical models of the perturbation and of the transfer function between the perturbation and the flux oscillations, was elaborated. This latter takes noise spectra from selected detectors as input and yields the perturbation position as output. The strength of the method lies in its potentially high spatial resolution, which is smaller than the typical distance between two adjacent LPRM detectors. The method was tested on simulated data, and then applied to the Forsmark measurements. The location of the disturbance, found by the algorithm, is in accordance with independent judgements for the case, and close to a position where an unseated assembly was found during refuelling. The purpose of this study was to develop and test the localisation method. To apply the

  9. Ground-based LiDAR application to characterize sea cliff instability processes along a densely populated coastline in Southern Italy

    Science.gov (United States)

    Esposito, Giuseppe; Semaan, Fouad; Salvini, Riccardo; Troise, Claudia; Somma, Renato; Matano, Fabio; Sacchi, Marco

    2017-04-01

    Sea cliff retreatment along the coastline of the Campi Flegrei volcanic area (Southern Italy) is becoming a threat for public and private structures due to the massive urbanization occurred in the last few decades. In this area, geological features of the outcropping rocks represent one of the most important factors conditioning the sea cliff retreatment. In fact, pyroclastic deposits formed by pumices, scoria, ashes and lapilli are arranged in weakly to moderately welded layers of variable thicknesses, resulting very erodible and prone to landslide processes. Available methods to evaluate topographic changes and retreat rates of sea cliffs include a variety of geomatic techniques, like terrestrial and aerial photogrammetry and LiDAR (Light Detection And Ranging). By means of such techniques, it is in fact possible to obtain high resolution topography of sea cliffs and perform multi-temporal change detection analysis. In this contribution, we present an application of Terrestrial Laser Scanning (TLS or ground-based LiDAR) aimed to identify and quantify instability processes acting along the Torrefumo coastal cliff, in the Campi Flegrei area. Specifically, we acquired a series of 3D point clouds on the years 2013 and 2016, and compared them through a cloud-to-cloud distance computation. Furthermore, a statistical analysis was applied to the change detection results. In this way, an inventory of the cliff failures occurred along the Torrefumo cliff in the 2013-2016 time span was created, as well as the spatial and volumetric distribution of these failures was evaluated. The volumetric analysis shows that large collapses occurred rarely, whereas the spatial analysis shows that the majority of failures occurred in the middle and upper parts of the cliff face. Results also show that both rock fall and surficial erosion processes contribute to the cliff retreatment, acting in turn according to the geological properties of the involved pyroclastic deposits. The presented

  10. Lower pressurization to increase BWR electric power under thermal hydraulic criteria

    International Nuclear Information System (INIS)

    Kataoka, Kazuyoshi; Chuman, Kazuto; Mizumachi, Wataru; Yoshioka, Ritsuo; Mori, Michitsugu; Horie, Akira; Machida, Yuzo

    1995-01-01

    Electric power output versus core size is one of the factors that determine the electricity generation costs of BWRs. The power output is roughly calculated from the thermal power of the BWR core and the thermal efficiency of the BWR turbine system. The thermal power is restricted by the reactor's thermal hydraulic criteria such as the maximum linear heat generation rate, the minimum critical power ratio, the pressure drop in the core and the feedwater temperature at the BWR inlet. The combination of a system pressure of approximately 5.5 MPa and a feedwater temperature of approximately 439 K offers the maximum electric power output for a BWR with 9 x 9 fuel bundles. The amount of electric power generated is about 9% more than that generated by a conventional BWR under the thermal hydraulic criteria. The electric power output increases as the system pressure and the feedwater temperature are decreased from the current design of 7.3 MPa and 488 K, respectively, because the increased critical power of the fuel bundles compensates for the lower thermal efficiency. (author)

  11. Risk-informed decision-making analysis for the electrical raceway fire barrier systems on a BWR-4 plant

    International Nuclear Information System (INIS)

    Wu, Ching-Hui; Lin, Tsu-Jen; Kao, Tsu-Mu; Chen, Chyn-Rong

    2003-01-01

    This paper describes a risk-informed decision-making approach used to resolve the fire barrier issue in a BWR-4 nuclear plant where Appendix R separation requirements cannot be met without installing additional fire protection features such as electrical raceway fire barrier system. The related risk measures in CDF (core damage frequency) and LERF (large early release frequency) of the fire barrier issue can be determined by calculating the difference in plant risks between various alternative cases and that met the requirement of the Appendix R. In some alternative cases, additional early-detection and fast-response fire suppression systems are suggested. In some other cases, cable re-routing of some improper layout of non-safety related cables are required. Sets of fire scenarios are re-evaluated more detailed by reviewing the cable damage impact for the BWR-4 plant. The fire hazard model, COMPBRM III-e, is used in this study and the dominant results in risk measures are benchmarked with the CFD code, FDS 2.0, to ensure that the risk impact of fire barrier is estimated accurately in the risk-informed decision making. The traditional deterministic qualitative methods, such as defense-in-depth, safety margin and post-fire safety shutdown capability are also proceeded. The value-impact analysis for proposed alternatives of fire wrapping required by Appendix R has been completed for technical basis of the exemption on Appendix R application. The outcome of the above analysis should be in compliance with the regulatory guidelines (RG) 1.174 and 1.189 for the applications in the risk-informed decision-making of the fire wrapping issues. (author)

  12. Model of oscillatory instability in vertically-homogeneous atmosphere

    Directory of Open Access Journals (Sweden)

    P. B. Rutkevich

    2009-02-01

    Full Text Available Existence and repeatability of tornadoes could be straightforwardly explained if there existed instability, responsible for their formation. However, it is well known that convection is the only instability in initially stable air, and the usual convective instability is not applicable for these phenomena. In the present paper we describe an instability in the atmosphere, which can be responsible for intense vortices. This instability appears in a fluid with Coriolis force and dissipation and has oscillatory behaviour, where the amplitude growth is accompanied by oscillations with frequency comparable to the growth rate of the instability. In the paper, both analytical analysis of the linear phase of the instability and nonlinear simulation of the developed stage of the air motion are addressed. This work was supported by the RFBR grant no. 09-05-00374-a.

  13. Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island design, Docket No. 50-447

    International Nuclear Information System (INIS)

    1983-04-01

    The Safety Evaluation Report for the application filed by General Electric Company for the Final Design Approval for the General Electric Standard Safety Analysis Report (GESSAR II FSAR) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report summarizes the results of the staff's safety review of the GESSAR II BWR/6 Nuclear Island Design. Subject to favorable resolution of items discussed in the Safety Evaluation Report, the staff concludes that the facilities referencing GESSAR II, subject to approval of the balance-of-plant design, can conform with the provisions of the Act and the regulations of the Nuclear Regulatory Commission

  14. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  15. Impact of metals recycling on a Swedish BWR decommissioning project

    International Nuclear Information System (INIS)

    Larsson, Arne; Lidar, Per; Hedin, Gunnar; Bergh, Niklas

    2014-01-01

    Decommissioning of nuclear power plants generates large volumes of radioactive or potentially contaminated metals. By proper management of the waste streams significant amounts can be free released and recycled either directly or after decontamination and melting. A significant part of the required work should be performed early in the process to make the project run smoothly without costly surprises and delays. A large portion of the clearance activities can be performed on-site. This on-site work should focus on the so called low-risk for contamination material. Other material can be decontaminated and released on site if schedule and the available facility areas so allow. It should be noted that the on-site decontamination and clearance activities can be a significant bottle neck for a decommissioning project. The availability of and access to a specialized metals recycling facility is an asset for a decommissioning project. This paper will describe the forecasted positive impact of a well-structured metals characterisation, categorisation and clearance process for a BWR plant decommissioning project. The paper is based on recent studies, performed projects and recent in-house development. (authors)

  16. Incorporation of an BWR evaporator concentrate in polyethylene

    International Nuclear Information System (INIS)

    Moriyama, N.; Dojiro, S.; Matsuzuru, H.

    1982-01-01

    The adaptability of polyethylene solidification method to an evaporator concentrate produced in BWR has been examined. The polyethylene product incorporating sodium sulfate as a main constituent of the concentrate was prepared using a batch-type melter. The product was tested with respect to homogeneity, density, mechanical property, water resistance and leachability, in order to evaluate the safety of the product. The product can incorporate the waste (sodium sulfate) up to 70 wt.% but it may be recommended through tests of water resistance and leachability that the maximum waste content in the product is 50 wt.%. The product with 50 wt.% waste has the density of 1.28 g/cm 3 , the compressive strength of 213 kgf/cm 2 , enough toughness, low leachability (diffusion coefficient of sodium is 10/sup -7/-10/sup ///31 /sup 6/ cm/sup 2//day) and excellent water resistance. This solidification method offers about 5 times larger volume reduction effect than a cement solidification does. 11 refs

  17. Technology developments for Japanese BWR MOX fuel utilization

    International Nuclear Information System (INIS)

    Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

    1997-01-01

    The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

  18. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  19. GOTHIC MODEL OF BWR SECONDARY CONTAINMENT DRAWDOWN ANALYSES

    International Nuclear Information System (INIS)

    Hansen, P.N.

    2004-01-01

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

  20. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  1. Beta and gamma dose calculations for PWR and BWR containments

    International Nuclear Information System (INIS)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 x 10 8 rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 x 10 8 rad equipment qualification test region. 8 refs., 23 figs., 12 tabs

  2. Controlling device for BWR type nuclear power plants

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1987-01-01

    Purpose: To enable to operate BWR type nuclear power plants while entirely drawing control rods upon stationary operation near the rated power. Method: Upon stationary operation of nuclear power plants near the rated power, an excess reactivity of nuclear fuels is controlled by entirely withdrawing control rods, and varying the feedwater enthalpy thereby changing the void amount. For this purpose, a feedwater heater is additionally disposed between a high pressure feedwater heater of a feedwater pipeway and a nuclear reactor pressure vessel, in which steams used for rising the temperature of the feedwater are introduced to the high temperature side of the high pressure feedwater heater and used again for the heating of the feedwater. In this case, the feedwater enthalpy can be increased approximately to the saturation enthalpy by adjusting such that about 30 % of the main steams are introduced to the high temperature side of the feedwater heater by a steam flow rate control valve, which enables to control the stationary operation without using control rods. (Horiuchi, T.)

  3. Calibration of the TVO spent BWR reference fuel assembly

    International Nuclear Information System (INIS)

    Tarvainen, M.; Baecklin, A.; Haakanson, A.

    1992-02-01

    In 1989 the Support Programmes of Finland (FSP) and Sweden (SSP) initiated a joint task to cross calibrate the burnup of the IAEA spent BWR reference fuel assembly at the TVO AFR storage facility (TVO KPA-STORE) in Finland. The reference assembly, kept separately under the IAEA seal, is used for verification measurements of spent fuel by GBUV method (SG-NDA-38). The cross calibration was performed by establishing a calibration curve, 244 Cm neutron rate versus burnup, using passive neutron assay (PNA) measurements. The declared burnup of the reference assembly was compared with the burnup value deduced from the calibration curve. A calibration line was also established by using the GBUV method with the aid of high resolution gamma ray spectrometry (HRGS). Normalization between the two different facilities was performed using sealed neutron and gamma calibration sources. The results of the passive neutron assay show consistency, better than 1 %, between the declared mean burnup of the reference assembly and the burnup deduced from the calibration curve. The corresponding consistency is within +-2 % for the HRGS measurements

  4. Natural heat transfer augmentation in passive advanced BWR plants

    International Nuclear Information System (INIS)

    Gamble, R.E.; Peterson, P.F.; Greif, R.

    2001-01-01

    In the European Simplified Boiling Water Reactor (ESBWR), the long-term post-accident containment pressure is determined by the combination of non condensable gas pressure and steam pressure in the wet well gas space. Since there are no active systems for heat removal in the wet well, energy transmitted to the wet well gas space, by a variety of means, must be removed by passive heat transfer to the walls and suppression pool (SP). The cold suppression pool located below the hotter gas space provides a stable configuration in which convection currents are suppressed thus limiting heat and mass transfer between the gas space and pool. However, heat transfer to the walls results in natural circulation currents that can augment the heat and mass transfer to the pool surface. Using a simplified model, parametric studies are carried out to show that augmentation of the order of magnitude expected can significantly impact the heat and mass transfer to the pool. Additionally a review of available literature in the area of augmentation and mixed convection of this type is presented and indicates the need for additional experimental work in order to develop adequate models for heat and mass transfer augmentation in the configuration of a BWR suppression pool. (author)

  5. BWR/5 Pressure-Suppression Pool Response during an SBO

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

    Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

  6. Experimental and numerical investigations of BWR fuel bundle inlet flow

    International Nuclear Information System (INIS)

    Hoashi, E; Morooka, S; Ishitori, T; Komita, H; Endo, T; Honda, H; Yamamoto, T; Kato, T; Kawamura, S

    2009-01-01

    We have been studying the mechanism of the flow pattern near the fuel bundle inlet of BWR using both flow visualization test and computational fluid dynamics (CFD) simulation. In the visualization test, both single- and multi-bundle test sections were used. The former test section includes only a corner orifice facing two support beams and the latter simulates 16 bundles surrounded by four beams. An observation window is set on the side of the walls imitating the support beams upstream of the orifices in both test sections. In the CFD simulation, as well as the visualization test, the single-bundle model is composed of one bundle with a corner orifice and the multi-bundle model is a 1/4 cut of the test section that includes 4 bundles with the following four orifices: a corner orifice facing the corner of the two neighboring support beams, a center orifice at the opposite side from the corner orifice, and two side orifices. Twin-vortices were observed just upstream of the corner orifice in the multi-bundle test as well as the single-bundle test. A single-vortex and a vortex filament were observed at the side orifice inlet and no vortex was observed at the center orifice. These flow patterns were also predicted in the CFD simulation using Reynolds Stress Model as a turbulent model and the results were in good agreement with the test results mentioned above. (author)

  7. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  8. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  9. Some studies for aseismic design of BWR power station

    International Nuclear Information System (INIS)

    Kasai, Hiroaki; Tanaka, Kihachiro; Takayanagi, Masaaki; Moriyama, Takeo; Kashimura, Yoshisada

    1978-01-01

    Nuclear power stations are constructed regarding the safety as the most important matter, and in Japan where earthquakes occur frequently, attention must be paid to the aseismatic characteristics in particular. BWR power generation facilities have high aseismatic performances on the basis of the experiences of construction and the design technology over many years. The tests and researches concerning the aseismatic characteristics are roughly divided into those of establishing reasonable design conditions, of improving the accuracy of the method of analyzing earthquake response, and of proving the maintenance of functions. Artificial earthquake waves are made by the method of correcting recorded waves or the method of sine wave synthesis. For the analysis of building-ground systems, the modeling method using finite element analysis has been studied. As for the researches on the aseismatic characteristics of core equipments, the vibration experiment on fuel assemblies and the proving test on the insertion of control rods were carried out. Especially in the proving test on the insertion of control rods, it was found that the stop of a nuclear reactor was able to be carried out without fail at the time of the earthquake far exceeding the maximum fuel amplitude estimated for the plants in Japan. Such results of researches are accumulated continuously and reflected to the design. (Kako, I.)

  10. MRI of shoulder instability

    Energy Technology Data Exchange (ETDEWEB)

    Steinbach, Lynne S. [University of California San Francisco, 505 Parnassus, Suite M392, San Francisco, CA 94143-0628 (United States)], E-mail: lynne.steinbach@radiology.ucsf.edu

    2008-10-15

    The most unstable joint in the body, the glenohumeral joint is subject to many insults including microinstability, subluxation and dislocation. During the last two decades, MRI has allowed for direct visualization of many of the lesions related to instability, aiding in diagnosis as well as therapeutic planning and follow-up. This article reviews the use of MRI for shoulder instability and describes the different types of lesions associated with this disorder. Topics include technical considerations, the use of MR arthrography, normal anatomy and variants, labral and glenohumeral ligament pathology, and osseous lesions related to instability.

  11. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor; Evaluacion del diseno radial de celdas de combustible en un ciclo de operacion de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez C, J.; Martin del Campo M, C. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jgco@ver.megared.net.mx

    2003-07-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  12. 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

  13. Design guideline to prevent the pipe rupture by radiolysis gases in BWR steam piping

    International Nuclear Information System (INIS)

    Inagaki, T.; Miyagawa, M.; Ota, T.; Sato, T.; Sakata, K.

    2009-01-01

    In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2007, TENPES published a revised edition of the guideline. This is the report of the revised edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent accumulation of radiolysis gas. (author)

  14. Analysis of a BWR direct cycle forced circulation power plants operation

    International Nuclear Information System (INIS)

    Andrade, G.G. de.

    1973-01-01

    First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

  15. 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.

  16. Hydrodynamic instabilities in miscible fluids

    Science.gov (United States)

    Truzzolillo, Domenico; Cipelletti, Luca

    2018-01-01

    Hydrodynamic instabilities in miscible fluids are ubiquitous, from natural phenomena up to geological scales, to industrial and technological applications, where they represent the only way to control and promote mixing at low Reynolds numbers, well below the transition from laminar to turbulent flow. As for immiscible fluids, the onset of hydrodynamic instabilities in miscible fluids is directly related to the physics of their interfaces. The focus of this review is therefore on the general mechanisms driving the growth of disturbances at the boundary between miscible fluids, under a variety of forcing conditions. In the absence of a regularizing mechanism, these disturbances would grow indefinitely. For immiscible fluids, interfacial tension provides such a regularizing mechanism, because of the energy cost associated to the creation of new interface by a growing disturbance. For miscible fluids, however, the very existence of interfacial stresses that mimic an effective surface tension is debated. Other mechanisms, however, may also be relevant, such as viscous dissipation. We shall review the stabilizing mechanisms that control the most common hydrodynamic instabilities, highlighting those cases for which the lack of an effective interfacial tension poses deep conceptual problems in the mathematical formulation of a linear stability analysis. Finally, we provide a short overview on the ongoing research on the effective, out of equilibrium interfacial tension between miscible fluids.

  17. Hydrodynamic instabilities in miscible fluids.

    Science.gov (United States)

    Truzzolillo, Domenico; Cipelletti, Luca

    2018-01-24

    Hydrodynamic instabilities in miscible fluids are ubiquitous, from natural phenomena up to geological scales, to industrial and technological applications, where they represent the only way to control and promote mixing at low Reynolds numbers, well below the transition from laminar to turbulent flow. As for immiscible fluids, the onset of hydrodynamic instabilities in miscible fluids is directly related to the physics of their interfaces. The focus of this review is therefore on the general mechanisms driving the growth of disturbances at the boundary between miscible fluids, under a variety of forcing conditions. In the absence of a regularizing mechanism, these disturbances would grow indefinitely. For immiscible fluids, interfacial tension provides such a regularizing mechanism, because of the energy cost associated to the creation of new interface by a growing disturbance. For miscible fluids, however, the very existence of interfacial stresses that mimic an effective surface tension is debated. Other mechanisms, however, may also be relevant, such as viscous dissipation. We shall review the stabilizing mechanisms that control the most common hydrodynamic instabilities, highlighting those cases for which the lack of an effective interfacial tension poses deep conceptual problems in the mathematical formulation of a linear stability analysis. Finally, we provide a short overview on the ongoing research on the effective, out of equilibrium interfacial tension between miscible fluids.

  18. Spondylolisthesis and Posterior Instability

    International Nuclear Information System (INIS)

    Niggemann, P.; Beyer, H.K.; Frey, H.; Grosskurth, D.; Simons, P.; Kuchta, J.

    2009-01-01

    We present the case of a patient with a spondylolisthesis of L5 on S1 due to spondylolysis at the level L5/S1. The vertebral slip was fixed and no anterior instability was found. Using functional magnetic resonance imaging (MRI) in an upright MRI scanner, posterior instability at the level of the spondylolytic defect of L5 was demonstrated. A structure, probably the hypertrophic ligament flava, arising from the spondylolytic defect was displaced toward the L5 nerve root, and a bilateral contact of the displaced structure with the L5 nerve root was shown in extension of the spine. To our knowledge, this is the first case described of posterior instability in patients with spondylolisthesis. The clinical implications of posterior instability are unknown; however, it is thought that this disorder is common and that it can only be diagnosed using upright MRI

  19. Spondylolisthesis and Posterior Instability

    Energy Technology Data Exchange (ETDEWEB)

    Niggemann, P.; Beyer, H.K.; Frey, H.; Grosskurth, D. (Privatpraxis fuer Upright MRT, Koeln (Germany)); Simons, P.; Kuchta, J. (Media Park Klinik, Koeln (Germany))

    2009-04-15

    We present the case of a patient with a spondylolisthesis of L5 on S1 due to spondylolysis at the level L5/S1. The vertebral slip was fixed and no anterior instability was found. Using functional magnetic resonance imaging (MRI) in an upright MRI scanner, posterior instability at the level of the spondylolytic defect of L5 was demonstrated. A structure, probably the hypertrophic ligament flava, arising from the spondylolytic defect was displaced toward the L5 nerve root, and a bilateral contact of the displaced structure with the L5 nerve root was shown in extension of the spine. To our knowledge, this is the first case described of posterior instability in patients with spondylolisthesis. The clinical implications of posterior instability are unknown; however, it is thought that this disorder is common and that it can only be diagnosed using upright MRI.

  20. Design of a redundant meteorological station for a BWR reactor; Diseno de una estacion meteorologica redundante para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, R.; Celis del Angel, L.; Bucio, F.; Rivero, T.; Palacios, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: ramses@nuclear.inin.mx

    2008-07-01

    In this work the design of a meteorological station for a reactor type BWR is proposed. Two independent channels of data acquisition that allow him to have a bigger readiness is exposed. It is incorporate sensors without mobile parts to measure speed, wind direction and pluvial precipitation. It also counts, with sensors of global solar radiation, net radiation, barometric pressure, relative humidity and ambient temperature; with them they are possible to be calculated, moreover, other variables as temperature differential, dew point and atmospheric stability. The sensors are placed on a tower to different heights and send their information (each second) to a local registration system, the one which in turn, it remits the data to the monitoring office so that a computer is linked with the system, display and management the information in real time and automatic way. The redundant structure allows that in the event of maintenance the data acquisition is not interrupted, even if the information is transferred to another place. In all the station sections it is used protocols of standard communication to allow that a great quantity of devices can be connected without major problem. The above-mentioned would allow to the operators in the control room to have reliable information during the whole time of the reactor operation. (Author)

  1. Streaming gravity mode instability

    International Nuclear Information System (INIS)

    Wang Shui.

    1989-05-01

    In this paper, we study the stability of a current sheet with a sheared flow in a gravitational field which is perpendicular to the magnetic field and plasma flow. This mixing mode caused by a combined role of the sheared flow and gravity is named the streaming gravity mode instability. The conditions of this mode instability are discussed for an ideal four-layer model in the incompressible limit. (author). 5 refs

  2. Report on the BWR owners group radiation protection/ALARA Committee

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, L.R. [Commonwealth Edison Co., Downers Grove, IL (United States)

    1995-03-01

    Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.

  3. Report on the BWR owners group radiation protection/ALARA Committee

    International Nuclear Information System (INIS)

    Aldrich, L.R.

    1995-01-01

    Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming open-quotes World Classclose quotes performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance

  4. 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

  5. BWR SFAT, gross-defect verification of spent BWR fuel. Final report on Task FIN A563 on the Finnish Support Programme to IAEA Safeguards including BWR SFAT User Manual

    International Nuclear Information System (INIS)

    Tarvainen, M.; Paakkunainen, M.; Tiitta, A.; Sarparanta, K.

    1994-04-01

    A measurement instrument called Spent Fuel Attribute Tester, SFAT, has been designed, fabricated and taken into use by the IAEA in gross defect verification of spent BWR fuel assemblies. The equipment consists of an underwater measurement head connected with cables to a control unit on the bridge of the fuel handling machine as well as to a PMCA for measurement of the gamma spectra. The BWR SFAT is optimized for the AFR interim storage, TVO KPA-STORE, of the TVO Power Company in Olkiluoto, Finland. It has a shape and it is moved like a fuel assembly using the fuel handling machine. No fuel movements are needed. Spent fuel specific radiation from the fission product 137 Cs at the gamma-ray energy of 662 keV is detected above the assemblies in the storage rack using a NaI(Tl) detector. In the design and in licensing the requirements of the IAEA, operator and the safety authority have been taken into account. The BWR SFAT allows modifications for other LWR fuel types with minor changes. The work has been carried out under the task FIN A 563 of the Finnish Support Programme to IAEA Safeguards. (orig.) (9 refs., 22 figs.)

  6. 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

  7. Automatic refueling platform and CRD remote handling device for BWR plant

    International Nuclear Information System (INIS)

    Kato, Hiroaki; Takagi, Kaoru

    1978-01-01

    In BWR plants, machines for replacing fuel assemblies and control rod drives are usually operated directly by personnel. An automatic refueling platform and a CRD remote handling device aiming at radiation exposure reduction and handling perfectness are described, which are already used in BWR plants. Automation of the former is achieved in transporting fuel assemblies between a reactor pressure vessel and a fuel storage pool, shuffling fuel assemblies in a reactor core and moving fuel assemblies in a fuel storage pool. In the latter, replacement of CRDs is nearly all performed remotely. (Mori, K.)

  8. A linear reactivity model approximation to predict BWR fuel assembly burnup

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Alonso V, G. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)]. e-mail: adl@nuclear.inin.mx

    2007-07-01

    A second order nodal method was derived for the one and a half group version of neutron diffusion theory in X-Y geometry. Based on the linear reactivity model, two reactivity relations are developed to treat fresh and depleted loading batches as a way to account for the spatial variation of moderator properties due to the substantial fraction of steam void present as well as the effect of the control rod positioning. Based on these ideas a program was developed to predict fuel batch burnup in a BWR operating cycle and fuel power sharing. The method was used to test two consecutive BWR operating fuel cycles. (Author)

  9. Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Aounallah, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2001-07-01

    This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

  10. The adiabatic instability on cosmology's dark side

    International Nuclear Information System (INIS)

    Bean, Rachel; Flanagan, Eanna E; Trodden, Mark

    2008-01-01

    We consider theories with a nontrivial coupling between the matter and dark energy sectors. We describe a small scale instability that can occur in such models when the coupling is strong compared to gravity, generalizing and correcting earlier treatments. The instability is characterized by a negative sound speed squared of an effective coupled dark matter/dark energy fluid. Our results are general, and applicable to a wide class of coupled models and provide a powerful, redshift-dependent tool, complementary to other constraints, with which to rule many of them out. A detailed analysis and applications to a range of models are presented in a longer companion paper

  11. Kinetic analyses on startup and shutdown chemistry of BWR plant

    International Nuclear Information System (INIS)

    Domae, Masafumi; Fujiwara, Kazutoshi; Inagaki, Hiromitsu

    2012-09-01

    During startup and shutdown of Boiling Water Reactor (BWR) plants, temperature and dissolved oxygen (DO) concentration of reactor water change in a wide range. The changes result in variation of conductivity and pH of the reactor water. It has been speculated that the water chemistry change is due to dissolution of the oxides on fuel claddings and structural materials. However, detailed mechanism is not known. In the present paper, trend of recent water chemistry in several BWR plants during startup and shutdown is presented. Conductivity and pH are convenient indication of coolant purity. We tried to clarify the mechanism of the change in the conductivity and the pH value during startup and shutdown, based on the water chemistry data measured. In the water chemistry data, change in chromate concentration and Ni 2+ concentration is rather large. It is assumed that change in the chromate concentration and the Ni 2+ concentration results in the time variation of the conductivity and the pH value. It is reasonable to consider that the increase in the chromate concentration and the Ni 2+ concentration is ascribed to dissolution of Cr oxides and Ni oxides, respectively. A model of dissolution of the Cr oxides and the Ni oxides is proposed. A concept of finite inventory of the Cr oxides and the Ni oxides in the coolant system is introduced. The model is as follows. Chromate is generated by oxidation of the Cr oxides and the Cr dissolution rate depends on the DO concentration. The dissolution rate of chromate is in proportion to DO concentration, the inventory of Cr and difference between solubility limit and the chromate concentration. On the other hand, Ni 2+ is formed by dissolution of the Ni oxides, and DO is not necessary in this process. The dissolution rate of Ni 2+ is in proportion to the inventory of Ni and difference between solubility limit and the Ni 2+ concentration. Coolant is continuously purified, and the chromate concentration and the Ni 2+ concentration

  12. Development and application of reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim K.H

    1999-04-01

    the paper presented here, we elaborate a modified and successful factorisation technique that takes the noise structure and the approximations performed into account. Due to an unseated fuel element, a local density wave oscillation appeared in the Swedish BWR Forsmark-1 in 1996. It is of great practical importance to locate the responsible element as soon as possible. We have developed a method for the localisation of such a thermohydraulic channel instability. The localisation method was tested on simulated data and then applied to a measurement taken in Forsmark-1 in early 1997. The results demonstrate the applicability and accuracy of the method. The noise present in the Ringhals PWRs has been investigated and in the course of this work, we have developed a new strategy towards the analysis of core barrel vibrations (CBM). This new approach unites the CBM analysis in the time and frequency domains. We have also developed and applied a new model for the determination of vibration properties such as the amplitude, amount of anisotropy and the preferred direction of the vibrations. This method is quantitative and suitable for long-term trend analysis of CBM vibrations 69 refs, 19 figs

  13. Development and application of reactor noise diagnostics

    International Nuclear Information System (INIS)

    Karlsson, Joakim K.H.

    1999-04-01

    presented here, we elaborate a modified and successful factorisation technique that takes the noise structure and the approximations performed into account. Due to an unseated fuel element, a local density wave oscillation appeared in the Swedish BWR Forsmark-1 in 1996. It is of great practical importance to locate the responsible element as soon as possible. We have developed a method for the localisation of such a thermohydraulic channel instability. The localisation method was tested on simulated data and then applied to a measurement taken in Forsmark-1 in early 1997. The results demonstrate the applicability and accuracy of the method. The noise present in the Ringhals PWRs has been investigated and in the course of this work, we have developed a new strategy towards the analysis of core barrel vibrations (CBM). This new approach unites the CBM analysis in the time and frequency domains. We have also developed and applied a new model for the determination of vibration properties such as the amplitude, amount of anisotropy and the preferred direction of the vibrations. This method is quantitative and suitable for long-term trend analysis of CBM vibrations

  14. Analysis of BWR high burnup fuel in LOCA conditions

    International Nuclear Information System (INIS)

    Garcia Sedano, Pablo; Dey Navarro, Jose Manuel; Gallego Cabezon, Ines; Orive Moreno, Raul

    2004-01-01

    High Burnup Fuel Behaviour has been growing in importance since middle 80's when pellet microstructure changes (rim effect) and cladding oxidation rates increase were observed. Later on, Cadarache reactivity tests revealed cladding integrity failures below safety limits. These phenomena, occurred at high burnup, stressed the necessity of having a wide experimental data base that would allow to dispose non-extrapolated data of material properties submitted to higher burnups than 40000 MWd/TM and data of new materials at the same time. One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MTU rod average burnup. The technical bases to support those high burnup levels are being developed. One of the licensing points of concern is the behaviour of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/TM and H.B. Robinson at 67 GWd/MTU. When the ANL tests have been finished, a conservative Peak Cladding Temperature/ Equivalent Cladding Reacted (PCT/ECR) limit will be determine from the residual ductility tests to be applied to the high burnup fuel. This makes necessary to determine the behaviour of the high burnup fuel in LOCA conditions and to determine the available safety margin. In licensing LOCA calculations, corresponding to present core designs and future core designs, the calculated PCT and ECR values as a function of the fuel burnup could be used to determine the relative severity of LOCA for the high burnup fuel. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations. (authors)

  15. Timing criteria for supplemental BWR emergency response equipment

    International Nuclear Information System (INIS)

    Bickel, John H.

    2015-01-01

    The Great Tohuku Earthquake and subsequent Tsunami represented a double failure event which destroyed offsite power connections to Fukushima-Daiichi site and then destroyed on-site electrical systems needed to run decay heat removal systems. The accident could have been mitigated had there been supplemental portable battery chargers, supplemental pumps, and in-place piping connections to provide alternate decay heat removal. In response to this event in the USA, two national response centers, one in Memphis, Tennessee, and another in Phoenix, Arizona, will begin operation. They will be able to dispatch supplemental emergency response equipment to any nuclear plant in the U.S. within 24 hours. In order to define requirements for supplemental nuclear power plant emergency response equipment maintained onsite vs. in a regional support center it is necessary to confirm: (a) the earliest time such equipment might be needed depending on the specific scenario, (b) the nominal time to move the equipment from a storage location either on-site or within the region of a nuclear power plant, and (c) the time required to connect in the supplemental equipment to use it. This paper describes an evaluation process for a BWR-4 with a Mark I Containment starting with: (a) severe accident simulation to define best estimate times available for recovery based on the specific scenario, (b) identify the key supplemental response equipment needed at specific times to accomplish recovery of key safety functions, and (c) evaluate what types of equipment should be warehoused on-site vs. in regional response centers. (authors)

  16. Liquid films and droplet deposition in a BWR fuel element

    International Nuclear Information System (INIS)

    Damsohn, M.

    2011-01-01

    In the upper part of boiling water reactors (BWR) the flow regime is dominated by a steam-water droplet flow with liquid films on the nuclear fuel rod, the so called (wispy) annular flow regime. The film thickness and liquid flow rate distribution around the fuel rod play an important role especially in regard to so called dryout, which is the main phenomenon limiting the thermal power of a fuel assembly. The deposition of droplets in the liquid film is important, because this process sustains the liquid film and delays dryout. Functional spacers with different vane shapes have been used in recent decades to enhance droplet deposition and thus create more favorable conditions for heat removal. In this thesis the behavior of liquid films and droplet deposition in the annular flow regime in BWR bundles is addressed by experiments in an adiabatic flow at nearly ambient pressure. The experimental setup consists of a vertical channel with the cross-section resembling a pair of neighboring subchannels of a fuel rod bundle. Within this double subchannel an annular flow is established with a gas-water mixture. The impact of functional spacers on the annular flow behavior is studied closely. Parameter variations comprise gas and liquid flow rates, gas density and spacer shape. The setup is instrumented with a newly developed liquid film sensor that measures the electrical conductance between electrodes flush to the wall with high temporal and spatial resolution. Advanced post-processing methods are used to investigate the dynamic behavior of liquid films and droplet deposition. The topic is also assessed numerically by means of single-phase Reynolds-Averaged-Navier-Stokes CFD simulations of the flow in the gas core. For this the commercial code STAR-CCM+ is used coupled with additional models for the liquid film distribution and droplet motion. The results of the experiments show that the liquid film is quite evenly distributed around the circumference of the fuel rods. The

  17. 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)

  18. Two-stream instabilities from the lower-hybrid frequency to the electron cyclotron frequency: application to the front of quasi-perpendicular shocks

    Science.gov (United States)

    Muschietti, Laurent; Lembège, Bertrand

    2017-09-01

    Quasi-perpendicular supercritical shocks are characterized by the presence of a magnetic foot due to the accumulation of a fraction of the incoming ions that is reflected by the shock front. There, three different plasma populations coexist (incoming ion core, reflected ion beam, electrons) and can excite various two-stream instabilities (TSIs) owing to their relative drifts. These instabilities represent local sources of turbulence with a wide frequency range extending from the lower hybrid to the electron cyclotron. Their linear features are analyzed by means of both a dispersion study and numerical PIC simulations. Three main types of TSI and correspondingly excited waves are identified: i. Oblique whistlers due to the (so-called fast) relative drift between reflected ions/electrons; the waves propagate toward upstream away from the shock front at a strongly oblique angle (θ ˜ 50°) to the ambient magnetic field Bo, have frequencies a few times the lower hybrid, and have wavelengths a fraction of the ion inertia length c/ωpi. ii. Quasi-perpendicular whistlers due to the (so-called slow) relative drift between incoming ions/electrons; the waves propagate toward the shock ramp at an angle θ a few degrees off 90°, have frequencies around the lower hybrid, and have wavelengths several times the electron inertia length c/ωpe. iii. Extended Bernstein waves which also propagate in the quasi-perpendicular domain, yet are due to the (so-called fast) relative drift between reflected ions/electrons; the instability is an extension of the electron cyclotron drift instability (normally strictly perpendicular and electrostatic) and produces waves with a magnetic component which have frequencies close to the electron cyclotron as well as wavelengths close to the electron gyroradius and which propagate toward upstream. Present results are compared with previous works in order to stress some features not previously analyzed and to define a more synthetic view of these TSIs.

  19. Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  20. Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2006-07-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U{sup 235} and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix

  1. 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

  2. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Johnsen, G.W.; Lellouche, G.S.

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 degrees F) has a 95-percentile uncertainty of 14.4 K (26 degrees F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 degrees F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 degrees F)

  3. Effect of injection rate on platinum deposition behaviour on stainless steel under simulated BWR conditions

    International Nuclear Information System (INIS)

    Grundler, P.V.; Ramar, A.; Abolhassani-Dadras, S.; Guenther-Leopold, I.; Kivel, N.; Ritter, S.; Karastoyanov, V.

    2012-09-01

    On-line NobleChem TM is a technology developed by General Electric-Hitachi to mitigate stress corrosion cracking (SCC) in reactor internals and recirculation pipes of boiling water reactors (BWRs) without the negative side-effects of the classical hydrogen water chemistry (HWC). With this technology a more efficient reduction of the electrochemical corrosion potential (ECP) is achieved by the injection of noble metals (e.g. Pt) into the feed water during power operation. Platinum is claimed to deposit as very fine metallic particles on all water-wetted surfaces and to stay electrocatalytic over long periods. A research project has been started at the Paul Scherrer Institut (PSI) investigating the deposition and distribution behaviour of Pt in BWRs. For this project, a sophisticated high-temperature water loop has been constructed, in which stainless steel coupon specimens are exposed to simulated BWR water. During the tests Pt solution is injected into the loop and Pt deposits on the specimens. The ECP of the specimens and other parameters such as dissolved O 2 and H 2 concentration are measured in-situ. Scanning and transmission electron microscopy techniques are used afterwards to characterise the Pt distribution and the size distribution on the oxide layer of the specimens. Additionally the specimens are analysed by Laser Ablation - Inductively Coupled Plasma - Mass Spectrometry to evaluate the Pt concentration on the specimens' oxide layers. The results from a set of experiments with varying Pt injection rates are presented. The obtained data set shows that higher injection rates yield larger Pt particles due to the higher concentration achieved and that the ECP decreases faster with increasing injection rates. The ECP reached is to some extend dependent on the total amount of Pt injected as it has been found also in the power plant Leibstadt (KKL) during the initial application. (authors)

  4. Fingerprints of dynamical instabilities

    International Nuclear Information System (INIS)

    Chomaz, Ph.; Colonna, M.; Guarnera, A.

    1993-01-01

    It is explained why any reduced descriptions, such as mean field approximation, are stochastic in nature. It is shown that the introduction of this stochastic dynamics leads to a predictive theory in a statistical sens whatever the individual trajectories are characterized by the occurrence of bifurcations, instabilities or phase transitions. Concerning nuclear matter, the spinodal instability is discussed. In such a critical situation, the possibility to replace the stochastic part of the collision integral in the Boltzmann-Langevin model by the numerical noise associated with the finite number of test particles in ordinary BUU treatment is studied. It is shown that the fingerprints of these instabilities are kept during the evolution because of the relatively long recombination time compared with the typical time scales imposed by the Coulomb repulsion and the possible collective expansion. (author) 5 refs., 12 figs

  5. Instability and star evolution

    International Nuclear Information System (INIS)

    Mirzoyan, L.V.

    1981-01-01

    The observational data are discussed which testify that the phenomena of dynamical instability of stars and stellar systems are definite manifestations of their evolution. The study of these phenomena has shown that the instability is a regular phase of stellar evolution. It has resulted in the recognition of the most important regularities of the process of star formation concerning its nature. This became possible due to the discovery in 1947 of stellar associations in our Galaxy. The results of the study of the dynamical instability of stellar associations contradict the predictions of classical hypothesis of stellar condensation. These data supplied a basis for a new hypothesis on the formation of stars and nebulae by the decay of superdense protostars [ru

  6. Causes of genome instability

    DEFF Research Database (Denmark)

    Langie, Sabine A S; Koppen, Gudrun; Desaulniers, Daniel

    2015-01-01

    , genome instability can be defined as an enhanced tendency for the genome to acquire mutations; ranging from changes to the nucleotide sequence to chromosomal gain, rearrangements or loss. This review raises the hypothesis that in addition to known human carcinogens, exposure to low dose of other......Genome instability is a prerequisite for the development of cancer. It occurs when genome maintenance systems fail to safeguard the genome's integrity, whether as a consequence of inherited defects or induced via exposure to environmental agents (chemicals, biological agents and radiation). Thus...... chemicals present in our modern society could contribute to carcinogenesis by indirectly affecting genome stability. The selected chemicals with their mechanisms of action proposed to indirectly contribute to genome instability are: heavy metals (DNA repair, epigenetic modification, DNA damage signaling...

  7. Manipulation of interfacial instabilities by using a soft, deformable ...

    Indian Academy of Sciences (India)

    Multilayer flows are oftensusceptible to interfacial instabilities caused due to jump in viscosity/elasticity across thefluid–fluid interface. It is frequently required to manipulate and control these interfacial instabilities in various applications such as coating processes or polymer coextrusion. We demonstrate here the possibility ...

  8. Filamentation instability of large-amplitude Alfven waves

    International Nuclear Information System (INIS)

    Kuo, S.P.; Whang, M.H.; Lee, M.C.

    1988-01-01

    An instability that leads to the filamentation of large-amplitude Alfven waves and gives rise to purely growing density and magnetic field fluctuations is studied. The dispersion relation of the instability is derived, from which the threshold conditions and the growth rates of the instability are analyzed quantitatively for applications to the solar wind plasma. We have examined their dependence on the filamentation spectrum, the plasma β, and the pump frequency and intensity for both right-hand and left-hand circularly polarized Alfven waves. The excitation of filamentation instability for certain cases of interest is discussed and compared with that of the parametric decay and modulation instability. The relevance of the proposed instability with some observations is discussed. copyright American Geophysical Union 1988

  9. Reducing BWR O and M costs through on-line performance monitoring

    International Nuclear Information System (INIS)

    Jonas, T.; Gross, R.; Logback, F.; Josyula, R.

    1995-01-01

    Competition in the electric power industry has placed significant emphasis on reducing operating and maintenance (O and M) costs at nuclear facilities. Therefore, on-line performance monitoring to locate power losses for boiling water reactor (BWR) plants is creating tremendous interest. In addition, the ability to automate activities such as data collection, analysis, and reporting increases the efficiency of plant engineers and gives them more time to concentrate on solving plant efficiency problems. This capability is now available with a unique software product called GEBOPS. GE Nuclear Energy, in conjunction with Joint Venture partner Black and Veatch, has undertaken development of the General Electric/Black and Veatch On-line Performance System (GEBOPS), an on-line performance monitoring system for BWR plants. The experience and expertise of GE Nuclear Energy with BWR plants, coupled with the proven on-line monitoring software development experience and capability of Black and Veatch, provide the foundation for a unique product which addresses the needs of today's BWR plants

  10. Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

    2016-05-15

    Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

  11. 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

  12. International comparison calculations for a BWR lattice with adjacent gadolinium pins

    International Nuclear Information System (INIS)

    Maeder, C.; Wydler, P.

    1984-09-01

    The results of burnup calculations for a simplified BWR fuel element with two adjacent gadolinium rods are presented and discussed. Ten complete solutions were contributed by Denmark, France, Italy (3), Japan (3), Switzerland and the UK. Partial results obtained from Poland and the USA are included in an Appendix. (Auth.)

  13. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

    1992-01-01

    This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  14. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

  15. Boron concentration evolution in the temporary curtains of a BWR reactor. Burcur code

    International Nuclear Information System (INIS)

    Cano Aguado, M.; Perlado Martin, J.M.; Minguez Torres, E.

    1977-01-01

    The theoretical model and the user's guide of the code Burcur is included. This code analyzes the burnable poison concentration of the temporary curtains as a function of time, for BWR reactors of the 7 x 7 design. The computing time being reasonably short, the number of burnup steps is as high as necessary.(author) [es

  16. Gravitational Instability of Cylindrical Viscoelastic Medium ...

    Indian Academy of Sciences (India)

    because of possible applications to various objects like white dwarf matter, inte- rior of heavy planets, atmosphere of neutron star and ultra cold neutral plasma. Prajapati & Chhajlani (2013) studied the linear self-gravitational instability of finitely conducting magnetized viscoelastic fluid using the GH model and discussed.

  17. LOSP-initiated event tree analysis for BWR

    International Nuclear Information System (INIS)

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

    1989-03-01

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

  18. Mobile crud and transportation of radioactivity in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H-P. [Studsvik Nuclear AB, Nykoping (Sweden); LTU, Div. of Chemical Engineering, Lulea (Sweden); Hagg, J. [Ringhals AB, Varobacka (Sweden)

    2010-07-01

    Mobile crud is here referred to as a generic term for all types of particles that occur in the reactor water in BWRs and that are able to carry radioactivity. Previous results in this on-going series of studies in Swedish BWRs suggest that there are particles of different origins and function. A share may come from fuel crud and others may come from detachment, precipitation and dissolution processes in different parts of the BWR primary system, as well as from other system parts, such as the turbine/condenser. In addition, crud particles in this sense may come from purely mechanical processes such as degradation of graphite containing parts of the control rod drives. Therefore, the overall aim was to evaluate which particles are responsible for the transportation and distribution of radioactivity and also to clarify the chemical conditions under which they are formed. Furthermore the aim was to draw conclusions about how the chemistry would be like in order to avoid or at least minimize the formation of radioactivity distributing particles. A specific objective has also been to look into the importance of particle size for spreading of radioactivity in the primary system. Different types of crud particles are likely to have different characteristics in terms of function associated with transportation of radioactivity. The fuel crud is radioactive from the source and other types of crud can via surface processes, co-precipitation and other chemical and mechanical processes potentially affect the distribution of radioactivity in the primary system. In order to predict how operating parameters (e.g. stable, full power operation and scram) and chemical parameters (NWC/HWC/Zn, etc.) will affect the activity build-up on the system surfaces, it is important to know how the different types of crud are affected by these and related parameters. Fuel crud fixed on cladding ring samples, as well as mobile crud from the reactor water captured on filters, were examined by

  19. Development and recent trend of disign of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Kani, Jiro

    1977-01-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation. (Wakatsuki, Y.)

  20. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

    2008-01-01

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  1. A new relativistic viscous hydrodynamics code and its application to the Kelvin-Helmholtz instability in high-energy heavy-ion collisions

    Science.gov (United States)

    Okamoto, Kazuhisa; Nonaka, Chiho

    2017-06-01

    We construct a new relativistic viscous hydrodynamics code optimized in the Milne coordinates. We split the conservation equations into an ideal part and a viscous part, using the Strang spitting method. In the code a Riemann solver based on the two-shock approximation is utilized for the ideal part and the Piecewise Exact Solution (PES) method is applied for the viscous part. We check the validity of our numerical calculations by comparing analytical solutions, the viscous Bjorken's flow and the Israel-Stewart theory in Gubser flow regime. Using the code, we discuss possible development of the Kelvin-Helmholtz instability in high-energy heavy-ion collisions.

  2. Buneman instability and Pierce instability in a collisionless bounded plasma

    International Nuclear Information System (INIS)

    Iizuka, Satoru; Saeki, Koichi; Sato, Noriyoshi; Hatta, Yoshisuke

    1983-01-01

    A systematic experiment is performed on the Buneman instability and the Pierce instability in a bounded plasma consisting of beam electrons and stationary ions. Current fluctuations are confirmed to be induced by the Buneman instability. On the other hand, the Pierce instability gives rise to a current limitation. The phenomena are well explained by Mikhailovskii's theory taking account of ion motion in a bounded plasma. (author)

  3. Neutrino beam plasma instability

    Indian Academy of Sciences (India)

    Abstract. We derive relativistic fluid set of equations for neutrinos and electrons from relativistic. Vlasov equations with Fermi weak interaction force. Using these fluid equations, we obtain a dis- persion relation describing neutrino beam plasma instability, which is little different from normal dispersion relation of streaming ...

  4. Elbow joint instability

    DEFF Research Database (Denmark)

    Olsen, Bo Sanderhoff; Henriksen, M G; Søjbjerg, Jens Ole

    1994-01-01

    The effect of simultaneous ulnar and radial collateral ligament division on the kinematics of the elbow joint is studied in a cadaveric model. Severance of the anterior part of the ulnar collateral ligament and the annular ligament led to significant elbow joint instability in valgus and varus...

  5. Structural and Material Instability

    DEFF Research Database (Denmark)

    Cifuentes, Gustavo Cifuentes

    This work is a small contribution to the general problem of structural and material instability. In this work, the main subject is the analysis of cracking and failure of structural elements made from quasi-brittle materials like concrete. The analysis is made using the finite element method. Three...

  6. Atlantooccipital instability. Case report.

    Science.gov (United States)

    Finney, H L; Roberts, T S

    1978-04-01

    A patient is presented who had neck pain and transient episodes of visual-field defect followed by unconsciousness when he hyperextended his neck. Radiographic examination revealed atlantooccipital instability on flexion and extension views. Treatment by fusion of the occiput to C-2 resulted in relief of symptoms.

  7. Tracking Code for Microwave Instability

    International Nuclear Information System (INIS)

    Heifets, S.; SLAC

    2006-01-01

    To study microwave instability the tracking code is developed. For bench marking, results are compared with Oide-Yokoya results [1] for broad-band Q = 1 impedance. Results hint to two possible mechanisms determining the threshold of instability

  8. Kinetic simulations of Rayleigh-Taylor instabilities

    International Nuclear Information System (INIS)

    Sagert, Irina; Bauer, Wolfgang; Colbry, Dirk; Howell, Jim; Staber, Alec; Strother, Terrance

    2014-01-01

    We report on an ongoing project to develop a large scale Direct Simulation Monte Carlo code. The code is primarily aimed towards applications in astrophysics such as simulations of core-collapse supernovae. It has been tested on shock wave phenomena in the continuum limit and for matter out of equilibrium. In the current work we focus on the study of fluid instabilities. Like shock waves these are routinely used as test-cases for hydrodynamic codes and are discussed to play an important role in the explosion mechanism of core-collapse supernovae. As a first test we study the evolution of a single-mode Rayleigh-Taylor instability at the interface of a light and a heavy fluid in the presence of a gravitational acceleration. To suppress small-wavelength instabilities caused by the irregularity in the separation layer we use a large particle mean free path. The latter leads to the development of a diffusion layer as particles propagate from one fluid into the other. For small amplitudes, when the instability is in the linear regime, we compare its position and shape to the analytic prediction. Despite the broadening of the fluid interface we see a good agreement with the analytic solution. At later times we observe the development of a mushroom like shape caused by secondary Kelvin-Helmholtz instabilities as seen in hydrodynamic simulations and consistent with experimental observations.

  9. Tidal instability in exoplanetary systems evolution

    Directory of Open Access Journals (Sweden)

    Le Gal P.

    2011-02-01

    Full Text Available A new element is proposed to play a role in the evolution of extrasolar planetary systems: the tidal (or elliptical instability. It comes from a parametric resonance and takes place in any rotating fluid whose streamlines are (even slightly elliptically deformed. Based on theoretical, experimental and numerical works, we estimate the growth rate of the instability for hot-jupiter systems, when the rotation period of the star is known. We present the physical process, its application to stars, and preliminary results obtained on a few dozen systems, summarized in the form of a stability diagram. Most of the systems are trapped in the so-called "forbidden zone", where the instability cannot grow. In some systems, the tidal instability is able to grow, at short timescales compared to the system evolution. Implications are discussed in the framework of misaligned transiting systems, as the rotational axis of the star would be unstable in systems where this elliptical instability grows.

  10. Inertioelastic Flow Instability at a Stagnation Point

    Science.gov (United States)

    Burshtein, Noa; Zografos, Konstantinos; Shen, Amy Q.; Poole, Robert J.; Haward, Simon J.

    2017-10-01

    A number of important industrial applications exploit the ability of small quantities of high molecular weight polymer to suppress instabilities that arise in the equivalent flow of Newtonian fluids, a particular example being turbulent drag reduction. However, it can be extremely difficult to probe exactly how the polymer acts to, e.g., modify the streamwise near-wall eddies in a fully turbulent flow. Using a novel cross-slot flow configuration, we exploit a flow instability in order to create and study a single steady-state streamwise vortex. By quantitative experiment, we show how the addition of small quantities (parts per million) of a flexible polymer to a Newtonian solvent dramatically affects both the onset conditions for this instability and the subsequent growth of the axial vorticity. Complementary numerical simulations with a finitely extensible nonlinear elastic dumbbell model show that these modifications are due to the growth of polymeric stress within specific regions of the flow domain. Our data fill a significant gap in the literature between the previously reported purely inertial and purely elastic flow regimes and provide a link between the two by showing how the instability mode is transformed as the fluid elasticity is varied. Our results and novel methods are relevant to understanding the mechanisms underlying industrial uses of weakly elastic fluids and also to understanding inertioelastic instabilities in more confined flows through channels with intersections and stagnation points.

  11. Multiphase Instabilities in Explosive Dispersal of Particles

    Science.gov (United States)

    Rollin, Bertrand; Ouellet, Frederick; Annamalai, Subramanian; Balachandar, S. ``Bala''

    2015-11-01

    Explosive dispersal of particles is a complex multiphase phenomenon that can be observed in volcanic eruptions or in engineering applications such as multiphase explosives. As the layer of particles moves outward at high speed, it undergoes complex interactions with the blast-wave structure following the reaction of the energetic material. Particularly in this work, we are interested in the multiphase flow instabilities related to Richmyer-Meshkov (RM) and Rayleigh-Taylor (RM) instabilities (in the gas phase and particulate phase), which take place as the particle layer disperses. These types of instabilities are known to depend on initial conditions for a relatively long time of their evolution. Using a Eulerian-Lagrangian approach, we study the growth of these instabilities and their dependence on initial conditions related to the particulate phase - namely, (i) particle size, (ii) initial distribution, and (iii) mass ratio (particles to explosive). Additional complexities associated with compaction of the layer of particles are avoided here by limiting the simulations to modest initial volume fraction of particles. A detailed analysis of the initial conditions and its effects on multiphase RM/RT-like instabilities in the context of an explosive dispersal of particles is presented. This work was supported by the U.S. Department of Energy, National Nuclear Security Administration, Advanced Simulation and Computing Program, as a Cooperative Agreement under the Predictive Science Academic Alliance Program, Contract No. DE-NA0002378.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  13. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

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

    Energy Technology Data Exchange (ETDEWEB)

    Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su' ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Riyana, EkaSapta [Nuclear Energy Regulatory Agency (BAPETEN) (Indonesia)

    2014-09-30

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

  15. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  16. A new relativistic viscous hydrodynamics code and its application to the Kelvin-Helmholtz instability in high-energy heavy-ion collisions

    Energy Technology Data Exchange (ETDEWEB)

    Okamoto, Kazuhisa [Nagoya University, Department of Physics, Nagoya (Japan); Nonaka, Chiho [Nagoya University, Department of Physics, Nagoya (Japan); Nagoya University, Kobayashi-Maskawa Institute for the Origin of Particles and the Universe (KMI), Nagoya (Japan); Duke University, Department of Physics, Durham, NC (United States)

    2017-06-15

    We construct a new relativistic viscous hydrodynamics code optimized in the Milne coordinates. We split the conservation equations into an ideal part and a viscous part, using the Strang spitting method. In the code a Riemann solver based on the two-shock approximation is utilized for the ideal part and the Piecewise Exact Solution (PES) method is applied for the viscous part. We check the validity of our numerical calculations by comparing analytical solutions, the viscous Bjorken's flow and the Israel-Stewart theory in Gubser flow regime. Using the code, we discuss possible development of the Kelvin-Helmholtz instability in high-energy heavy-ion collisions. (orig.)

  17. Instability and internet design

    Directory of Open Access Journals (Sweden)

    Sandra Braman

    2016-09-01

    Full Text Available Instability - unpredictable but constant change in one’s environment and the means with which one deals with it - has replaced convergence as the focal problem for telecommunications policy in general and internet policy in particular. Those who designed what we now call the internet during the first decade of the effort (1969-1979, who in essence served simultaneously as its policy-makers, developed techniques for coping with instability of value for network designers today and for those involved with any kind of large-scale sociotechnical infrastructure. Analysis of the technical document series that was medium for and record of that design process reveals coping techniques that began with defining the problem and went on to include conceptual labour, social practices, and technical approaches.

  18. Orphans and political instability.

    Science.gov (United States)

    Breuning, Marijke; Ishiyama, John

    2011-01-01

    This study investigates the security implications of growing orphan populations, particularly in Sub-Saharan Africa. Little has been written about the security implications of this especially vulnerable group of children. Are growing orphan populations associated with increases in political instability as has been suggested? Using data from several sources, we employ regression analysis to test whether Sub-Saharan African countries with larger proportions of orphans and those with increasing orphan populations experience higher rates of political instability. We find that the increase in the orphan population is related to an increasing incidence of civil conflict, but do not find a similar relationship for the proportion of orphans. In addition, we find that the causes of orphanhood matter. We conclude that increases in orphan populations (rather than simple proportions) are destabilizing. We suggest possible avenues for mediating the security risks posed by growing orphan populations.

  19. State Instability and Terrorism

    Science.gov (United States)

    2010-01-01

    security, heterosexual relationships and goods and services. Although prisons constitute a very controlled social structure to begin with, it is...closely although the NBRM model fits slightly better. The Vuong test preferred the ZINB model over the NBRM (p<.0000). This suggests that I ought to...then, it is preferred because it allows me to examine the effects of instability within-country only. This is the purest test of the 109 concept

  20. Instability of gravitating sphalerons

    International Nuclear Information System (INIS)

    Boschung, P.; Brodbeck, O.; Moser, F.; Straumann, N.; Volkov, M.

    1994-01-01

    We prove the instability of the gravitating regular sphaleron solutions of the SU(2) Einstein-Yang-Mills-Higgs system with a Higgs doublet, by studying the frequency spectrum of a class of radial perturbations. With the help of a variational principle we show that there exist always unstable modes. Our method has the advantage that no detailed knowledge of the equilibrium solution is required. It does, however, not directly apply to black holes

  1. Relativistic centrifugal instability

    Science.gov (United States)

    Gourgouliatos, Konstantinos N.; Komissarov, Serguei S.

    2018-03-01

    Near the central engine, many astrophysical jets are expected to rotate about their axis. Further out they are expected to go through the processes of reconfinement and recollimation. In both these cases, the flow streams along a concave surface and hence, it is subject to the centrifugal force. It is well known that such flows may experience the centrifugal instability (CFI), to which there are many laboratory examples. The recent computer simulations of relativistic jets from active galactic nuclei undergoing the process of reconfinement show that in such jets CFI may dominate over the Kelvin-Helmholtz instability associated with velocity shear (Gourgouliatos & Komissarov). In this letter, we generalize the Rayleigh criterion for CFI in rotating fluids to relativistic flows using a heuristic analysis. We also present the results of computer simulations which support our analytic criterion for the case of an interface separating two uniformly rotating cylindrical flows. We discuss the difference between CFI and the Rayleigh-Taylor instability in flows with curved streamlines.

  2. Instability of warped discs

    Science.gov (United States)

    Doǧan, S.; Nixon, C. J.; King, A. R.; Pringle, J. E.

    2018-01-01

    Accretion discs are generally warped. If a warp in a disc is too large, the disc can `break' apart into two or more distinct planes, with only tenuous connections between them. Further if an initially planar disc is subject to a strong differential precession, then it can be torn apart into discrete annuli that precess effectively independently. In previous investigations, torque-balance formulae have been used to predict where and when the disc breaks into distinct parts. In this work, focusing on discs with Keplerian rotation and where the shearing motions driving the radial communication of the warp are damped locally by turbulence (the `diffusive' regime), we investigate the stability of warped discs to determine the precise criterion for an isolated warped disc to break. We find and solve the dispersion relation, which in general yields three roots. We provide a comprehensive analysis of this viscous-warp instability and the emergent growth rates and their dependence on disc parameters. The physics of the instability can be understood as a combination of (1) a term which would generally encapsulate the classical Lightman-Eardley instability in planar discs (given by ∂(νΣ)/∂Σ condition acting on the diffusion of the warp amplitude given in simplified form by ∂(ν2|ψ|)/∂|ψ| < 0. We discuss our findings in the context of discs with an imposed precession, and comment on the implications for different astrophysical systems.

  3. Analysis of results of AZTRAN and AZKIND codes for a BWR; Analisis de resultados de los codigos AZTRAN y AZKIND para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M., E-mail: gbo729@yahoo.com.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

  4. Development of a computerized operator support system for BWR power plant

    International Nuclear Information System (INIS)

    Monta, K.; Sekimizu, K.; Sato, N.; Araki, T.; Mori, N.

    1985-01-01

    A computerized operator support system for BWR power plant has been developed since 1980 supported by the Japanese government. The main functions of the systems are post trip operational guidance, disturbance analysis, standby system management, operational margin monitoring and control rod operational guidance. The former two functions aim at protection against incidents during operation of nuclear power plants and the latter three functions aim at their prevention. As the final stage of the development, these functions are combined with the plant supervision function and are organized as an advanced man-machine interface for BWR power plant. During the above process, operator task analyses are performed to enable synthesis of these support functions for right fit to operator tasks and to realize a hierarchical structure for CRT displays for right fit to operators cognitive needs. (author)

  5. Study of the possibility of power uprate of a simplified BWR

    International Nuclear Information System (INIS)

    Kim, H.T.; Yang, A.I.; Tabata, H.; Montani, M.; Yoshioka, Y.

    1995-01-01

    Stability and critical power analyses were performed to study the possibility of power uprate for a large natural circulation BWR beyond 1000 MWe based on the thermal hydraulic performance requirement. The natural circulation flow required to satisfy the stability and critical power design criteria and the chimney height necessary to provide the natural circulation flow were determined by performing a parameter sensitivity study. Based on the study results, relationship between power level, fuel type, and size of reactor pressure vessel (diameter and height) was established. This paper presents the study results showing that power uprate beyond 1000 MWe for a large natural circulation BWR is quite feasible from the thermal hydraulic performance standpoint. (author)

  6. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH Univ. of Applied Sciences, Deggendorf (Germany)

    2014-07-01

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation programme was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment with integrated pressure suppression system. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The main target was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. (orig.)

  7. Analysis of radiological consequences in a typical BWR with a mark-II containment

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro

    2003-01-01

    INS/NUPEC in Japan has been carrying out the Level 3 PSA program. In the program, the MACCS2 code has been extensively applied to analyze radiological consequences for typical BWR and PWR plants in Japan. The present study deals with analysis of effects of the AMs, which were implemented by industries, on radiological consequence for a typical BWR with a Mark-II containment. In the present study, source terms and their frequencies of source terms were used based on results of Level 2 PSA taking into account AM countermeasures. Radiological consequences were presented with dose risks (Sv/ry), which were multiplied doses (Sv) by containment damage frequencies (/ry), and timing of radionuclides release to the environment. The results of the present study indicated that the dose risks became negligible in most cases taking AM countermeasures and evacuations. (author)

  8. Aggressive chemical decontamination tests on small valves from the Garigliano BWR

    International Nuclear Information System (INIS)

    Bregani, F.

    1990-01-01

    In order to check the effectiveness of direct chemical decontamination on small and complex components, usually considered for storage without decontamination because of the small amount, some tests were performed on the DECO experimental loop. Four small stainless steel valves from the primary system of the Garigliano BWR were decontaminated using mainly aggressive chemicals such as HC1, HF, HNO 3 and their mixtures. On two valves, before the treatment with aggressive chemicals, a step with soft chemical (oxalic and citric acid mixture) was performed in order to see whether a softening action enhances the following aggressive decontamination. Moreover, in order to increase as much as possible the decontamination effectiveness, a decontamination process using ultrasounds jointly with aggressive chemicals was investigated. After an intensive laboratory testing programme, two smaller stainless steel valves from the primary system of the Garigliano BWR were decontaminated using ultrasounds in aggressive chemical solutions

  9. Assessment of the fracture toughness of irradiated stainless steel for BWR core shrouds

    International Nuclear Information System (INIS)

    Carter, R.G.; Gamble, R.M.

    2002-01-01

    Data from previously performed experiments were collected and evaluated to determine the relationship between fracture toughness and neutron fluence for conditions representative of BWR core shrouds. This relationship together with EPFM (elastic-plastic fracture mechanics) analysis methods similar to those in Appendix K of Section XI of the ASME Code were used to compute margin against failure as a function of neutron fluence for postulated cracks in BWR core shrouds. The results indicate that EPFM analyses can be used for flaw evaluation of core shrouds at fluence levels less than 3.10 21 n/cm 2 (E > 1 MeV). At fluence levels equal to or greater than 3.10 21 n/cm 2 , LEFM (linear-elastic fracture mechanics) analyses should be used with K Ic = 55 MPa-(m) 0.5 . (authors)

  10. 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)

  11. Decontamination as a precursor to decommissioning. Status report Task 2: process evaluation. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Divine, J.R.; Woodruff, E.M.; McPartland, S.A.; Zima, G.E.

    1983-05-01

    As part of the US Nuclear Regulatory Commission's program to reduce occupational exposure and waste volumes, the Pacific Northwest Laboratory is studying decontamination as a precursor to decommissioning. Eleven processes or solvents were examined for their behavior in decontaminating BWR carbon steel samples. The solvents included NS-1, a proprietary solvent of Dow Chemical Corporation, designed for BWR use, and AP-Citrox, a well-known, two-step process designed for PWR stainless steel; it was used to provide a reference for later comparison to other systems and processes. The decontamination factors observed in the tests performed in a small laboratory scale recirculating loop ranged from about 1 (no effect) to 222 (about 99.6% of the initial activity removed. Coordinated corrosion measurements were made using twelve chemical solvents and eight metal alloys found in a range of reactor types.

  12. Investigations on the thermal-hydraulics of a natural circulation cooled BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kok, H.V.; Hagen, T.H.J.J. van der; Mudde, R.F. [Delft Univ. of Technology (Netherlands)

    1995-09-01

    A scaled natural circulation loop facility has been built after the Dodewaard Boiling Water Reactor, which is the only operating natural circulation cooled BWR in the world. The loop comprises one fuel assembly, a riser with a downcomer and a condenser with a cooling system. Freon-12 is used as a scaling liquid. This paper reports on the first measurements done with this facility. Quantities like the circulation flow, carry-under and the void-fraction have been measured as a function of power, pressure, liquid level, riser length, condensate temperature and friction factors. The behavior of the circulation flow can be understood by considering the driving force. Special attention has been paid to the carry-under, which has been shown to have a very important impact on the dynamics of a natural circulation cooled BWR.

  13. 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

  14. The distribution of the activity in the primary system of a BWR

    International Nuclear Information System (INIS)

    Westerberg, S.; Hermansson, H.-P.

    1980-12-01

    A model for the distribution of activity is presented. The model is based upon an earlier mathematical model which has been modified and applied to the calculation of activity of the primary water system and to the buildup of activity in ion exchange materials. Calculations have been performed for Co-60. A certain agreement of the calculations and the data from Swedish BWR plants has been obtained. (G.B.)

  15. BWR Mark I pressure suppression study: effect of downcomer fill level on the vertical load function

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1977-02-01

    The investigation reported forms a part of the BWR Mark I Pressure Suppression Experiment Program and is one of a series of small scale studies designed to evaluate limited aspects of the pool dynamics phenomena prior to conduct of the 1/5 scale air test series. Presented is an experimental study of the effect of downcomer fill level (DFL) on the vertical load function

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  17. Product Evaluation Task Force Phase Two report for BWR/PWR dissolver wastes

    International Nuclear Information System (INIS)

    Francis, A.J.

    1990-01-01

    It has been proposed that all Intermediate Level Wastes arising at Sellafield should be encapsulated prior to ultimate disposal. The Product Evaluation Task Force (PETF) was set up to investigate possible encapsulants and to produce an adequate data base to justify the preferred matrices. This report details the work carried out, under Phase 2 of the Product Evaluation Task Force programme, on BWR/PWR Dissolver Wastes. Three possible types of encapsulants for BWR/PWR Dissolver Wastes:- Inorganic cements, Polymer cements and Polymers are evaluated using the Kepner Tregoe decision analysis technique. This technique provides a methodology for scoring and ranking alternative options and evaluating any risks associated with an option. The analysis shows that for all four stages of waste management operations ie Storage, Transport, handling and emplacement, Disposal and Process, cement matrices are considerably superior to other potential matrices. A matrix, consisting of three parts Blast Furnace Slag (BFS) to one part Ordinary Portland Cement (OPC), is recommended for Phase 3 studies on BWR/PWR Dissolver Wastes. (author)

  18. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  19. BWR control blade/channel box interaction models for SCDAP/RELAP5

    International Nuclear Information System (INIS)

    Griffin, F.P.

    1993-01-01

    The core of a boiling water reactor (BWR) consists of an array of fuel assemblies with cross-shaped control blades located between these assemblies. Each fuel assembly consists of a fuel rod bundle surrounded by a Zircaloy channel box. Each control blade consists of small stainless steel absorber tubes filled with B 4 C powder surrounded by a stainless steel blade sheath. Under severe accident conditions, material interactions between the B 4 C, stainless steel, and Zircaloy would have a significant impact on the melting and subsequent relocation of the control blade and channel box structures. This paper describes a new BWR control blade/channel box model for the SCDAP/RELAP5 severe accident analysis code that includes the effects of these material interactions. The phenomena represented by this model and the modeling techniques are derived from ORNL analyses of the BWR severe fuel damage experiments. Two examples of the operation of this new model within SCDAP/RELAP5 are provided

  20. Development of a coordinated control system for BWR nuclear power plant and HVDC transmission system

    International Nuclear Information System (INIS)

    Ishikawa, M.; Hara, T.; Hirayama, K.; Sekiya, K.

    1986-01-01

    The combined use of dc and ac transmissions or so-called hybrid transmission was under study, employing both dc and ac systems to enable stable transmission of 10,000 MW of electric power generated by the BWR nuclear plant, scheduled to be built about 800 km away from the center of the load. It was thus necessary to develop a hybrid power transmission control system, the hybrid power transmission system consisting of a high voltage dc transmission system (HVDC) and an ultrahigh ac transmission system (UHVAC). It was also necessary to develop a control system for HVDC transmission which protects the BWR nuclear power plant from being influenced by any change in transmission mode that occurs as a result of faults on the UHVAC side when the entire power of the BWR plant is being sent by the HVDC transmission. This paper clarifies the requirements for the HVDC system control during hybrid transmission and also during dc transmission. The control method that satisfies these requirements was studied to develop a control algorithm

  1. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  2. Analysis CFD for the hydrogen transport in the primary containment of a BWR

    International Nuclear Information System (INIS)

    Jimenez P, D. A.; Del Valle G, E.; Gomez T, A. M.

    2014-10-01

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  3. Thermomechanical analysis of a fuel rod in a BWR reactor using the FUELSIM code; Analisis termomecanico de una barra de combustible de un reactor BWR utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, IPN, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico, D. F. (Mexico); Ortiz V, J.; Araiza M, E. [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rapaca78@yahoo.com.mx

    2009-10-15

    The thermomechanical behaviour of a fuel rod exposed to irradiation is a complex process in which are coupled great quantity of interrelated physical-chemical phenomena, for that analysis of rod performance in the core of a nuclear power reactor is realized generally with computation codes that integrate several phenomena expected during the time life of fuel rod in the core. An application of this type of thermomechanical codes is to predict, inside certain reliability margin, the design parameters that would be required to adjust, in order to get a better economy or rod performance, for a systematic approach to the fuel design optimization. FUELSIM is a thermomechanical code based on the models of FRAPCON code, which was developed under auspice of Nuclear Regulatory Commission of USA. FUELSIM allows iterative calculations like part of its programming structure, allowing search of extreme cases of behaviour, probabilistic analysis (or statistical), parametric analysis (or sensibility) and also can include as entrance data to the uncertainties associated with production data, code parameters and associated models. In this work is reported a first analysis of thermomechanical performance of a typical fuel rod used in a BWR 5/6. Results of maximum temperatures are presented in the fuel center and of axial deformation, for the 10 axial nodes in that the active longitude of fuel rod was divided. (Author)

  4. Calculation of the neutron flux and fluence in the covering of the nucleus and the vessel of a BWR; Calculo del flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor nuclear BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: evalle@esfm.ipn.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    One of the main objectives related with the safety in any nuclear power plant, including the nuclear power plant of Laguna Verde, is to guarantee the structural integrity of the pressure vessel of the reactor. To identify and quantifying the damage caused be neutron irradiation in the vessel of any nuclear reactor, is necessary to know as much the neutron flux as the fluence that it has been receiving during their time of operation life, since the observables damages by means of tests mechanics are products of micro-structural effects, induced by neutron irradiation, therefore, is important the study and prediction of the neutron flux to have a better knowledge of the damage that are receiving these materials. In our calculation the code DORT was used, which solves the transport equation in discreet coordinates and in two dimensions (x-y, r-{theta} and r-z), in accord to the regulator guide, it requires to make and approach of the neutron flux in three dimensions by means of the Synthesis Method. Whit this method is possible to achieve a representation of the flux in 3D combining or synthesizing the calculated fluxes by DORT code in r-{theta}, r-z and r. In this work the application of the Synthesis Method is presented, according to the Regulator Guide 1.190, to determine the fluxes 3D in the interns of a BWR using three different space meshes. (Author)

  5. Optimization of axial enrichment distribution for BWR fuels using scoping libraries and block coordinate descent method

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

    Highlights: • An optimization method for axial enrichment distribution in a BWR fuel was developed. • Block coordinate descent method is employed to search for optimal solution. • Scoping libraries are used to reduce computational effort. • Optimization search space consists of enrichment difference parameters. • Capability of the method to find optimal solution is demonstrated. - Abstract: An optimization method has been developed to search for the optimal axial enrichment distribution in a fuel assembly for a boiling water reactor core. The optimization method features: (1) employing the block coordinate descent method to find the optimal solution in the space of enrichment difference parameters, (2) using scoping libraries to reduce the amount of CASMO-4 calculation, and (3) integrating a core critical constraint into the objective function that is used to quantify the quality of an axial enrichment design. The objective function consists of the weighted sum of core parameters such as shutdown margin and critical power ratio. The core parameters are evaluated by using SIMULATE-3, and the cross section data required for the SIMULATE-3 calculation are generated by using CASMO-4 and scoping libraries. The application of the method to a 4-segment fuel design (with the highest allowable segment enrichment relaxed to 5%) demonstrated that the method can obtain an axial enrichment design with improved thermal limit ratios and objective function value while satisfying the core design constraints and core critical requirement through the use of an objective function. The use of scoping libraries effectively reduced the number of CASMO-4 calculation, from 85 to 24, in the 4-segment optimization case. An exhausted search was performed to examine the capability of the method in finding the optimal solution for a 4-segment fuel design. The results show that the method found a solution very close to the optimum obtained by the exhausted search. The number of

  6. 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

  7. Radiation Induced Genomic Instability

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, William F.

    2011-03-01

    Radiation induced genomic instability can be observed in the progeny of irradiated cells multiple generations after irradiation of parental cells. The phenotype is well established both in vivo (Morgan 2003) and in vitro (Morgan 2003), and may be critical in radiation carcinogenesis (Little 2000, Huang et al. 2003). Instability can be induced by both the deposition of energy in irradiated cells as well as by signals transmitted by irradiated (targeted) cells to non-irradiated (non-targeted) cells (Kadhim et al. 1992, Lorimore et al. 1998). Thus both targeted and non-targeted cells can pass on the legacy of radiation to their progeny. However the radiation induced events and cellular processes that respond to both targeted and non-targeted radiation effects that lead to the unstable phenotype remain elusive. The cell system we have used to study radiation induced genomic instability utilizes human hamster GM10115 cells. These cells have a single copy of human chromosome 4 in a background of hamster chromosomes. Instability is evaluated in the clonal progeny of irradiated cells and a clone is considered unstable if it contains three or more metaphase sub-populations involving unique rearrangements of the human chromosome (Marder and Morgan 1993). Many of these unstable clones have been maintained in culture for many years and have been extensively characterized. As initially described by Clutton et al., (Clutton et al. 1996) many of our unstable clones exhibit persistently elevated levels of reactive oxygen species (Limoli et al. 2003), which appear to be due dysfunctional mitochondria (Kim et al. 2006, Kim et al. 2006). Interestingly, but perhaps not surprisingly, our unstable clones do not demonstrate a “mutator phenotype” (Limoli et al. 1997), but they do continue to rearrange their genomes for many years. The limiting factor with this system is the target – the human chromosome. While some clones demonstrate amplification of this chromosome and thus lend

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  9. Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2002-07-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

  11. Study of microwave instability

    International Nuclear Information System (INIS)

    Heifets, S.; Chao, A.

    1996-12-01

    The microwave instability is usually described by linearizing Vlasov equation in the angle-action variables I, φ and assuming that the interaction of azimuthal harmonics ρ n (I) of the distribution function ρ is weak. The argument implied here is that the Hamiltonian flow smears out particles over invariant tori characterized by the action variables, and the remaining azimuthal dependence of the distribution function is small. Indeed, such an approach successfully describes bunch spectrum and the threshold of the microwave instability. However, recently there have been interesting observations of bunch centroid and bunch shape oscillations above instability threshold at LEP and the damping ring at SLAC. There are also indicates that the oscillations sometimes occur in localized region in the longitudinal coordinate instead of affecting the entire longitudinal distribution as one expects by an action-angle analysis. In this paper the authors describe an alternative approach to the problem of bunch stability using decomposition of the Fokker-Plank equation in the system of nonlinear equations for the moments of the distribution function. In particular, this approach allows them to avoid the conventional action-angle decomposition. The physical quantities they are interested in, the moments, are expressed in the Cartesian z - δ phase space. To close the infinite hierarchy of moments equations, the authors assume that higher order correlations are small. Although both the action-angle and the Cartesian languages must be equivalent before truncation, they may have different speed of convergence depending on the problem being studied. It is hoped that Cartesian expansion approach would converge faster for the cases corresponding to those observed recently above threshold. The recent experimental observations made them interested in it again. This note is a progress report of their work

  12. Instability of warped discs

    Science.gov (United States)

    Doǧan, S.; Nixon, C. J.; King, A. R.; Pringle, J. E.

    2018-05-01

    Accretion discs are generally warped. If a warp in a disc is too large, the disc can `break' apart into two or more distinct planes, with only tenuous connections between them. Further, if an initially planar disc is subject to a strong differential precession, then it can be torn apart into discrete annuli that precess effectively independently. In previous investigations, torque-balance formulae have been used to predict where and when the disc breaks into distinct parts. In this work, focusing on discs with Keplerian rotation and where the shearing motions driving the radial communication of the warp are damped locally by turbulence (the `diffusive' regime), we investigate the stability of warped discs to determine the precise criterion for an isolated warped disc to break. We find and solve the dispersion relation, which, in general, yields three roots. We provide a comprehensive analysis of this viscous-warp instability and the emergent growth rates and their dependence on disc parameters. The physics of the instability can be understood as a combination of (1) a term that would generally encapsulate the classical Lightman-Eardley instability in planar discs (given by ∂(νΣ)/∂Σ < 0) but is here modified by the warp to include ∂(ν1|ψ|)/∂|ψ| < 0, and (2) a similar condition acting on the diffusion of the warp amplitude given in simplified form by ∂(ν2|ψ|)/∂|ψ| < 0. We discuss our findings in the context of discs with an imposed precession, and comment on the implications for different astrophysical systems.

  13. High beta tokamak instabilities

    International Nuclear Information System (INIS)

    Bateman, G.

    1977-01-01

    Theoretical predictions using the ideal MHD model indicable that large-scale ballooning modes should appear when the average beta is raised about 1 to 2% in present-day tokamak geometries or 5 to 10% in more optimized geometries. The onset of instability is predicted to be sudden and the behavior of ballooning modes to be strikingly different from the saw-tooth and Mirnov oscillations experimentally observed at low beta. Conditions close to the predicted onset were achieved in ORMAK with no noticeable change in plasma behavior. Experiments are planned for the ISX tokamak to test the beta limit. 15 references, 3 figures

  14. Evaporation and Antievaporation Instabilities

    Directory of Open Access Journals (Sweden)

    Andrea Addazi

    2017-10-01

    Full Text Available We review (antievaporation phenomena within the context of quantum gravity and extended theories of gravity. The (antievaporation effect is an instability of the black hole horizon discovered in many different scenarios: quantum dilaton-gravity, f ( R -gravity, f ( T -gravity, string-inspired black holes, and brane-world cosmology. Evaporating and antievaporating black holes seem to have completely different thermodynamical features compared to standard semiclassical black holes. The purpose of this review is to provide an introduction to conceptual and technical aspects of (antievaporation effects, while discussing problems that are still open.

  15. Semiconductor Lasers Stability, Instability and Chaos

    CERN Document Server

    Ohtsubo, Junji

    2013-01-01

    This third edition of “Semiconductor Lasers, Stability, Instability and Chaos” was significantly extended.  In the previous edition, the dynamics and characteristics of chaos in semiconductor lasers after the introduction of the fundamental theory of laser chaos and chaotic dynamics induced by self-optical feedback and optical injection was discussed. Semiconductor lasers with new device structures, such as vertical-cavity surface-emitting lasers and broad-area semiconductor lasers, are interesting devices from the viewpoint of chaotic dynamics since they essentially involve chaotic dynamics even in their free-running oscillations. These topics are also treated with respect to the new developments in the current edition. Also the control of such instabilities and chaos control are critical issues for applications. Another interesting and important issue of semiconductor laser chaos in this third edition is chaos synchronization between two lasers and the application to optical secure communication. One o...

  16. Standard Technical Specifications General Electric plants, BWR/4:Bases (Sections 3.4-3.10). Volume 3, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the specifications for all chapters and sections of the improved STS. Volume 2 contains he Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. This document, Volume 3, contains the Bases for Sections 3.4-3.10 of the improved STS

  17. Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

    Science.gov (United States)

    Tanaka, Ken-ichi; Ueno, Jun

    2017-09-01

    Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR). Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

  18. Resonant Drag Instabilities in protoplanetary disks: the streaming instability and new, faster-growing instabilities

    Science.gov (United States)

    Squire, Jonathan; Hopkins, Philip F.

    2018-04-01

    We identify and study a number of new, rapidly growing instabilities of dust grains in protoplanetary disks, which may be important for planetesimal formation. The study is based on the recognition that dust-gas mixtures are generically unstable to a Resonant Drag Instability (RDI), whenever the gas, absent dust, supports undamped linear modes. We show that the "streaming instability" is an RDI associated with epicyclic oscillations; this provides simple interpretations for its mechanisms and accurate analytic expressions for its growth rates and fastest-growing wavelengths. We extend this analysis to more general dust streaming motions and other waves, including buoyancy and magnetohydrodynamic oscillations, finding various new instabilities. Most importantly, we identify the disk "settling instability," which occurs as dust settles vertically into the midplane of a rotating disk. For small grains, this instability grows many orders of magnitude faster than the standard streaming instability, with a growth rate that is independent of grain size. Growth timescales for realistic dust-to-gas ratios are comparable to the disk orbital period, and the characteristic wavelengths are more than an order of magnitude larger than the streaming instability (allowing the instability to concentrate larger masses). This suggests that in the process of settling, dust will band into rings then filaments or clumps, potentially seeding dust traps, high-metallicity regions that in turn seed the streaming instability, or even overdensities that coagulate or directly collapse to planetesimals.

  19. Curvature-Induced Instabilities of Shells

    Science.gov (United States)

    Pezzulla, Matteo; Stoop, Norbert; Steranka, Mark P.; Bade, Abdikhalaq J.; Holmes, Douglas P.

    2018-01-01

    Induced by proteins within the cell membrane or by differential growth, heating, or swelling, spontaneous curvatures can drastically affect the morphology of thin bodies and induce mechanical instabilities. Yet, the interaction of spontaneous curvature and geometric frustration in curved shells remains poorly understood. Via a combination of precision experiments on elastomeric spherical shells, simulations, and theory, we show how a spontaneous curvature induces a rotational symmetry-breaking buckling as well as a snapping instability reminiscent of the Venus fly trap closure mechanism. The instabilities, and their dependence on geometry, are rationalized by reducing the spontaneous curvature to an effective mechanical load. This formulation reveals a combined pressurelike term in the bulk and a torquelike term in the boundary, allowing scaling predictions for the instabilities that are in excellent agreement with experiments and simulations. Moreover, the effective pressure analogy suggests a curvature-induced subcritical buckling in closed shells. We determine the critical buckling curvature via a linear stability analysis that accounts for the combination of residual membrane and bending stresses. The prominent role of geometry in our findings suggests the applicability of the results over a wide range of scales.

  20. Efficiency versus instability in plasma accelerators

    Science.gov (United States)

    Lebedev, Valeri; Burov, Alexey; Nagaitsev, Sergei

    2017-12-01

    Plasma wakefield acceleration is one of the main technologies being developed for future high-energy colliders. Potentially, it can create a cost-effective path to the highest possible energies for e+e- or γ -γ colliders and produce a profound effect on the developments for high-energy physics. Acceleration in a blowout regime, where all plasma electrons are swept away from the axis, is presently considered to be the primary choice for beam acceleration. In this paper, we derive a universal efficiency-instability relation, between the power efficiency and the key instability parameter of the trailing bunch for beam acceleration in the blowout regime. We also show that the suppression of instability in the trailing bunch can be achieved through Balakin-Novokhatsky-Smirnov damping by the introduction of a beam energy variation along the bunch. Unfortunately, in the high-efficiency regime, the required energy variation is quite high and is not presently compatible with collider-quality beams. We would like to stress that the development of the instability imposes a fundamental limitation on the acceleration efficiency, and it is unclear how it could be overcome for high-luminosity linear colliders. With minor modifications, the considered limitation on the power efficiency is applicable to other types of acceleration.

  1. Atlantooccipital instability in Down syndrome.

    Science.gov (United States)

    Rosenbaum, D M; Blumhagen, J D; King, H A

    1986-06-01

    Atlantooccipital instability is an uncommon and usually lethal result of major trauma. Nontraumatic cases of instability at this site are rare. We report the findings in two children with Down syndrome who have striking atlantooccipital subluxation demonstrated on flexion and extension radiographs. Since there is increasing radiologic evaluation of cervical spine stability in patients with Down syndrome who wish to participate in athletic activities, the status of the atlantooccipital joint needs careful assessment, especially after cervical fusion for C1-C2 instability.

  2. Clinical application of computer-aided design-rapid prototyping in C1-C2 operation techniques for complex atlantoaxial instability.

    Science.gov (United States)

    Yang, Jin-Cheng; Ma, Xiang-Yang; Xia, Hong; Wu, Zeng-Hui; Ai, Fu-Zhi; Zhang, Kai; Yin, Qing-Shui

    2014-06-01

    This is a retrospective clinical study. To evaluate the clinical efficacy of computer-aided design-rapid prototyping (CAD-RP) techniques in surgical treatments for atlantoaxial instability (AAI). The complexity of the upper cervical anatomic structures makes the procedures for the treatment of AAI particularly challenging for surgeons. The present study represents a series of C1-C2 surgery for AAI aided by CAD-RP. A total of 49 patients (21 men and 28 women) with AAI were treated in our department. According to the use of the CAD-RP technique, the patients were divided into RP group and No RP group. Preoperative CT scans of the upper cervical spine were performed for each patient. For the RP group, physical RP models of the upper cervical spine were manufactured from the 3-dimensional CT data and were used for intraoperative guidance. Personalized surgeries were performed for each case of the 2 groups. The screw malposition rate, frequency of using intraoperative fluoroscopy, operation time, blood loss, and improvement of neurological function were compared between the 2 groups. The mean follow-up duration was 32 months (range, 24-50 mo). The operations were successfully performed in 48 cases expect for 1 case in the No RP group. A total of 204 screws were placed. The intraoperative fluoroscopy frequency and operation time were significantly lower in the RP group than that in the No RP group in both posterior and anterior approaches, whereas the screw malposition rate showed no difference between the 2 groups for both approaches. After the operation, 48 cases achieved satisfactory decompression of the cervical cord and repositioning of the atlantoaxial spine. During follow-up, 47 cases presented improvements in the spinal nerve function within 2 years. CAD-RP techniques have significant benefits for surgeons providing personalized treatments for AAI, especially cases with complicated deformities.

  3. Calculations for a BWR Lattice with Adjacent Gadolinium Pins Using the Monte Carlo Cell Code Serpent v.1.1.7

    Directory of Open Access Journals (Sweden)

    Diego Ferraro

    2011-01-01

    Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.

  4. Suppression of electromechanical instability in fiber-reinforced dielectric elastomers

    Directory of Open Access Journals (Sweden)

    Rui Xiao

    2016-03-01

    Full Text Available The electromechanical instability of dielectric elastomers has been a major challenge for the application of this class of active materials. In this work, we demonstrate that dielectric elastomers filled with soft fiber can suppress the electromechanical instability and achieve large deformation. Specifically, we developed a constitutive model to describe the dielectric and mechanical behaviors of fiber-reinforced elastomers. The model was applied to study the influence of stiffness, nonlinearity properties and the distribution of fiber on the instability of dielectric membrane under an electric field. The results show that there exists an optimal fiber distribution condition to achieve the maximum deformation before failure.

  5. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume I. RELAP4/MOD5 description. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    RELAP4 is a computer program written in FORTRAN IV for the digital computer analysis of nuclear reactors and related systems. It is primarily applied in the study of system transient response to postulated perturbations such as coolant loop rupture, circulation pump failure, power excursions, etc. The program was written to be used for water-cooled (PWR and BWR) reactors and can be used for scale models such as LOFT and SEMISCALE. Additional versatility extends its usefulness to related applications, such as ice condenser and containment subcompartment analysis. Specific options are available for reflood (FLOOD) analysis and for the NRC Evaluation Model.

  6. Investigation of natural circulation instability and transients in passively safe novel modular reactor

    Science.gov (United States)

    Shi, Shanbin

    The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

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

    International Nuclear Information System (INIS)

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-01-01

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

  8. Instabilities in electromagnetic quasilevitation.

    Science.gov (United States)

    Spragg, Kirk; Letout, Sebastien; Ernst, R; Sneyd, Alfred; Fautrelle, Yves

    2014-05-01

    We investigate free-surface instabilities occurring in various industrial processes involving liquid metal. Of particular interest is the behavior of the free surface of a pool of liquid metal when it is submitted to an alternating magnetic field. Experimentally, we study the effect of a vertical alternating medium-frequency magnetic field on an initially circular pool. We observe various types of behavior according to magnetic field amplitude, e.g., axisymmetric deformations, azimuthal mode structures, slow radial oscillation of the pool perimeter, and random rotation of the pool around its center. Drop rotation could be attributed to nonsymmetric shape deformations. The effect of oxidation leads to drastic changes in pool behavior. The experimental results are then compared to a linear stability analysis of the free surface of a circular liquid drop.

  9. Posterolateral elbow joint instability

    DEFF Research Database (Denmark)

    Olsen, Bo Sanderhoff; Søjbjerg, Jens Ole; Nielsen, K K

    1998-01-01

    Thirty-five osteoligamentous elbows were included in a study on the kinematics of posterolateral elbow joint instability during the pivot shift test (PST) before and after separate ligament cuttings in the lateral collateral ligament complex (LCLC). Division of the annular ligament or the lateral...... ulnar collateral ligament caused no laxity during the PST. Division of the lateral collateral ligament caused maximal laxity of 4 degrees and 23 degrees during forced PST in valgus and external rotation (supination), respectively. Cutting of the LCLC at the ulnar or the humeral insertion was necessary...... for any PST stressed elbow joint laxity to occur. Total division of the LCLC induced a maximal laxity of 7.9 degrees and 37 degrees during forced PST in valgus and external rotation (supination), respectively. This study suggests the lateral collateral ligament to be the primary soft tissue constraint...

  10. Cinerama sickness and postural instability

    NARCIS (Netherlands)

    Bos, J.E.; Ledegang, W.D.; Lubeck, A.J.A.; Stins, J.F.

    2013-01-01

    Motion sickness symptoms and increased postural instability induced by motion pictures have been reported in a laboratory, but not in a real cinema. We, therefore, carried out an observational study recording sickness severity and postural instability in 19 subjects before, immediately and 45 min

  11. Cohabitation and Children's Family Instability

    Science.gov (United States)

    Kelly Raley, R.; Wildsmith, Elizabeth

    2004-01-01

    This study estimates how much children's family instability is missed when we do not count transitions into and out of cohabitation, and examines early life course trajectories of children to see whether children who experience maternal cohabitation face more family instability than children who do not. Using data from the 1995 National Survey of…

  12. Nonlinear theory of drift instability

    International Nuclear Information System (INIS)

    Hatori, T.

    1981-01-01

    In this chapter, we review recent works on the analytical and numerical analysis for the nonlinear evolution of drift instabilities. Only the case of a coherent wave is considered. Contributions to the turbulence theory for drift instabilities are already presented in Chapter 4. (author)

  13. Internal rotor friction instability

    Science.gov (United States)

    Walton, J.; Artiles, A.; Lund, J.; Dill, J.; Zorzi, E.

    1990-01-01

    The analytical developments and experimental investigations performed in assessing the effect of internal friction on rotor systems dynamic performance are documented. Analytical component models for axial splines, Curvic splines, and interference fit joints commonly found in modern high speed turbomachinery were developed. Rotor systems operating above a bending critical speed were shown to exhibit unstable subsynchronous vibrations at the first natural frequency. The effect of speed, bearing stiffness, joint stiffness, external damping, torque, and coefficient of friction, was evaluated. Testing included material coefficient of friction evaluations, component joint quantity and form of damping determinations, and rotordynamic stability assessments. Under conditions similar to those in the SSME turbopumps, material interfaces experienced a coefficient of friction of approx. 0.2 for lubricated and 0.8 for unlubricated conditions. The damping observed in the component joints displayed nearly linear behavior with increasing amplitude. Thus, the measured damping, as a function of amplitude, is not represented by either linear or Coulomb friction damper models. Rotordynamic testing of an axial spline joint under 5000 in.-lb of static torque, demonstrated the presence of an extremely severe instability when the rotor was operated above its first flexible natural frequency. The presence of this instability was predicted by nonlinear rotordynamic time-transient analysis using the nonlinear component model developed under this program. Corresponding rotordynamic testing of a shaft with an interference fit joint demonstrated the presence of subsynchronous vibrations at the first natural frequency. While subsynchronous vibrations were observed, they were bounded and significantly lower in amplitude than the synchronous vibrations.

  14. Dynamical Instability and Soliton Concept

    International Nuclear Information System (INIS)

    Kartavenko, V.G.

    1994-01-01

    The problem of dynamical instability and clustering (stable fragments formation) in a breakup of excited nuclear systems are considered from the points of view of the soliton concept. It is noted that the volume (spinodal) instability can be associated with nonlinear terms, and the surface (Rayleigh-Taylor type) instability, with the dispersion terms in the evolution equations. The spinodal instability and the Rayleigh-Taylor instability may compensate each other and lead to stable quasi-soliton type objects. The simple analytical model is presented to illustrate this physical picture. The time evolution of an initially compressed cold nuclear system is analysed in the framework of the inverse mean-field method. It is demonstrated that the nonlinearity and dispersion terms of the evolution equations can lead to clusterization in the final channel. 8 p

  15. Instability of ties in compression

    DEFF Research Database (Denmark)

    Buch-Hansen, Thomas Cornelius

    2013-01-01

    exact instability solutions are complex to derive, not to mention the extra complexity introducing dimensional instability from the temperature gradients. Using an inverse variable substitution and comparing an exact theory with an analytical instability solution a method to design tie...... the temperature gradient between the outer and the inner wall, which results in critical increase of the bending moments in the ties. Since the ties are loaded by combined compression and moment forces, the loadbearing capacity is derived from instability equilibrium equations. Most of them are iterative, since......-connectors in cavity walls was developed. The method takes into account constraint conditions limiting the free length of the wall tie, and the instability in case of pure compression which gives an optimal load bearing capacity. The model is illustrated with examples from praxis....

  16. Beam instability Workshop - plenary sessions

    International Nuclear Information System (INIS)

    2001-01-01

    The purpose of this workshop was to provide a review of the mechanisms of limiting beam instabilities, their cures, including feedback, and beam measurement for synchrotron radiation light sources. 12 plenary sessions took place whose titles are: 1) challenging brilliance and lifetime issues with increasing currents; 2) limiting instabilities in multibunch; 3) experience from high currents in B factories; 4) longitudinal dynamics in high intensity/bunch; 5) Transverse instabilities for high intensity/bunch; 6) working group introduction from ESRF experience; 7) impedance modelling: simulations, minimization; 8) report on the broadband impedance measurements and modelling workshop; 9) feedback systems for synchrotron light sources; 10) beam instabilities diagnostics; 11) harmonic cavities: the pros and cons; and 12) experimental study of fast beam-ion instabilities at PLS. This document gathers the 12 articles that were presented during these sessions

  17. Electron/electron acoustic instability

    International Nuclear Information System (INIS)

    Gary, S.P.

    1987-01-01

    The electron acoustic wave becomes a normal mode of an unmagnetized collisionless plasma in the presence of two electron components with similar densities, but strongly disparate temperatures. The characteristic frequency of this mode is the plasma frequency of the cooler electron component. If these two electron components have a relative drift speed several times the thermal speed of the cooler component, the electron/electron acoustic instability may arise. This paper describes the parametric dependences of the threshold drift speed and maximum growth rate of this instability, and compares these with the same properties of the electron/ion acoustic instability. Under the condition of zero current, the electron/ion acoustic instability typically has the lower threshold drift speed, so that observation of the electron/electron acoustic instability is a strong indication of the presence of an electrical current in the plasma

  18. Water chemistry control and decontamination experience with TEPCO BWR`s and the measures planned for the future

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, N.; Miyamaru, K. [Tokyo Electric Power Co. (Japan)

    1995-03-01

    The new TEPCO BWR`s are capable of having the occupational radiation exposure controlled successfully at a low level by selecting low cobalt steel, using corrosion-resistant steel, employing dual condensate polishing systems, and controlling Ni/Fe ratio during operation. The occupational radiation exposure of the old BWR`s, on the other hand, remains high though reduced substantially through the use of low cobalt replacement steel and the partial addition of a filter in the condensate polishing system. Currently under review is the overall decontamination procedure for the old BWR`s to find out to measures needed to reduce the amount of crud that is and has been carried over into the nuclear reactor. The current status of decontamination is reported below.

  19. Obtention control bars patterns for a BWR using Tabo search; Obtencion de patrones de barras de control para un BWR usando busqueda Tabu

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, A.; Ortiz, J.J.; Alonso, G. [Instituto Nacional de Investigaciones Nucleares, Km. 36.5 Carretera Mexico-Toluca, Ocoyoacac, Estado de Mexico 52045 (Mexico); Morales, L.B. [UNAM, IIMAS, Ciudad Universitaria, D. F. 04510 (Mexico); Valle, E. del [IPN, ESFM, Unidad Profesional ' Adolfo Lopez Mateos' , Col. Lindavista 07738, D. F. (Mexico)]. e-mail: jacm@nuclear.inin.mx

    2004-07-01

    The obtained results when implementing the technique of tabu search, for to optimize patterns of control bars in a BWR type reactor, using the CM-PRESTO code are presented. The patterns of control bars were obtained for the designs of fuel reloads obtained in a previous work, using the same technique. The obtained results correspond to a cycle of 18 months using 112 fresh fuels enriched at the 3.53 of U-235. The used technique of tabu search, prohibits recently visited movements, in the position that correspond to the axial positions of the control bars, additionally the tiempo{sub t}abu matrix is used for to manage a size of variable tabu list and the objective function is punished with the frequency of the forbidden movements. The obtained patterns of control bars improve the longitude of the cycle with regard to the reference values and they complete the restrictions of safety. (Author)

  20. Analysis of assemblies exchange in the core of a reactor BWR; Analisis del intercambio de ensambles en el nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    The performance of the core of a boiling water reactor (BWR) was evaluated when two assemblies are exchanged during the fuel reload in erroneous way. All with the purpose of analyzing the value of the neutrons effective multiplication factor and the thermal limits for an exchange of assemblies. In their realization the mentioned study was based in a transition cycle of the Unit 1 of the nuclear power plant of Laguna Verde. The obtained results demonstrate that when carrying out an exchange between two fuel assemblies in erroneous way, with regard to the original reload, the changes in the neutrons effective multiplication factor do not present a serious problem, unless the exchange has been carried out among a very burnt assembly with one fresh, where this last is taken to the periphery. (Author)

  1. Optimization of fuel reloads for a BWR using the ant colony system; Optimizacion de recargas de combustible para un BWR usando el sistema de colonia de hormigas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J. [Universidad Autonoma del Estado de Mexico, Facultad de Ingenieria, Cerro de Coatepec s/n, Ciudad Universitaria, 50110 Toluca, Estado de Mexico (Mexico); Ortiz S, J. J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jaime.es.jaime@gmail.com

    2009-10-15

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  2. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  3. Study of transient rod extraction failure without RBM in a BWR; Estudio del transitorio error de extraccion de barra sin RBM en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

  4. Study of transient turbine shot without bypass in a BWR; Estudio del transitorio disparo de turbina sin bypass en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

  5. Semiconductor Lasers Stability, Instability and Chaos

    CERN Document Server

    Ohtsubo, Junji

    2008-01-01

    This monograph describes fascinating recent progress in the field of chaos, stability and instability of semiconductor lasers. Applications and future prospects are discussed in detail. The book emphasizes the various dynamics induced in semiconductor lasers by optical and electronic feedback, optical injection, and injection current modulation. Recent results of both theoretical and experimental investigations are presented. Demonstrating applications of semiconductor laser chaos, control and noise, Semiconductor Lasers describes suppression and chaotic secure communications. For those who are interested in optics but not familiar with nonlinear systems, a brief introduction to chaos analysis is presented.

  6. Thermal instability in the interstellar medium

    Directory of Open Access Journals (Sweden)

    J. Ghanbari

    2000-06-01

    Full Text Available   This study demonstrates how thermal structures in the interstellar medium can emerge as a result of thermal instability. For a two-dimensional case, the steady state thermal structures was investigeted and it was shown that a large class of solutions exist. For a one –dimensional case the conductivity was found to be negligible. The effects of to cal cooling on the thermal instability were explored in some depth. In this case analytical results for time-dependent cooling function were presented, too. We studied nonlinear wave phenomena in thermal fluid systems, with a particular emphasis on presenting analytical results. When conductivity is proportional to temperature, the beliavior of thermal waves is soliton like. For slow thermal waves, approximate analytical results were presented. Extensions of this work are discussed briefly, together with possible astrophysical applications.

  7. Modeling elastic instabilities in nematic elastomers

    Science.gov (United States)

    Mbanga, Badel L.; Ye, Fangfu; Selinger, Jonathan V.; Selinger, Robin L. B.

    2010-11-01

    Liquid crystal elastomers are cross-linked polymer networks covalently bonded with liquid crystal mesogens. In the nematic phase, due to strong coupling between mechanical strain and orientational order, these materials display strain-induced instabilities associated with formation and evolution of orientational domains. Using a three-dimensional finite element elastodynamics simulation, we investigate one such instability, the onset of stripe formation in a monodomain film stretched along an axis perpendicular to the nematic director. In our simulation, we observe the formation of striped domains with alternating director rotation. This model allows us to explore the fundamental physics governing dynamic mechanical response of nematic elastomers and also provides a potentially useful computational tool for engineering device applications.

  8. Longwave instabilities and patterns in fluids

    CERN Document Server

    Shklyaev, Sergey

    2017-01-01

    This book summarizes the main advances in the field of nonlinear evolution and pattern formation caused by longwave instabilities in fluids. It will allow readers to master the multiscale asymptotic methods and become familiar with applications of these methods in a variety of physical problems.  Longwave instabilities are inherent to a variety of systems in fluid dynamics, geophysics, electrodynamics, biophysics, and many others. The techniques of the derivation of longwave amplitude equations, as well as the analysis of numerous nonlinear equations, are discussed throughout. This book will be of value to researchers and graduate students in applied mathematics, physics, and engineering, in particular within the fields of fluid mechanics, heat and mass transfer theory, and nonlinear dynamics. .

  9. Early Prevention Method for Power Systems Instability

    DEFF Research Database (Denmark)

    Dmitrova, Evgenia

    a given operational conditions and predicts the resulting stability margins for the new steady state, while avoiding time consuming time domain simulations. The method has been validated on the Western Danish power system model, containing 464 buses. The case study of aperiodic small signal angular....... The predicted effect of the suggested countermeasure application is in a good agreement with the results obtained by RMS dynamic simulation. Developed method enables adaptive preventive control for near real-time stability maintenance. The achieved results are opening promising perspective for power system’s......In the scope of this work, a method capable of fast identification of the proper countermeasure, that prevents emerging instability, has been developed. The focus is placed on the prevention of aperiodic small signal angular instability by means of manipulations applied to load nodes (nodes...

  10. Estimation of dose rate around the spent control rods of a BWR; Estimacion de la rapidez de dosis alrededor de las barras de control gastadas de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cancino P, G.

    2016-10-01

    The energy can come from fossil renewable sources (solar (natural gas, oil), wind, hydro, tidal, geothermal, biomass, bio energy and nuclear. Nuclear power can be obtained by fission reactions and fusion (still under investigation) atomic nuclei. Fission, is a partition of a very heavy nucleus (Uranium 235, for example) into two lighter nuclei. Much of the world's electric power is generated from the energy released by fission processes. In a nuclear power reactor, light water as the BWR, there are many important elements that allow safe driving operation, one of them are the elements or control systems, the burnable poison or neutron absorber inherently allow control power reactor. The control rods, which consist mostly of stainless steel and absorbing elements (such as boron carbide, hafnium, cadmium, among others) of thermal neutrons is able to initiate, regulate or stop the reactor power. These, due to the use of depleted burned or absorbing material and therefore reach their lifespan, which can be 15 years or have other values depending on the manufacturer. Control rods worn should be removed, stored or confined in expressly places. Precisely at this stage arises the importance of knowing their radiological condition to manipulate safely and without incident to the people health responsible for conducting these proceedings state arises. This thesis consists in the estimation of the dose rate in spent control rod made of boron carbide, from a typical BWR reactor. It will be estimated by direct radiation measurements with measurement equipment for radiotherapy ionization chamber, in six spent control rods, which were taken at different reactor operating cycles and are in a spent fuel pool. Using bracket electromechanical and electronic equipment for positioning and lifting equipment for radiation measurement around the control rod in the axial and radial arrangement for proper scanning. Finally will be presented a graphic corresponding to the dose

  11. Functional Instability of the Ankle Joint: Etiopathogenesis

    Directory of Open Access Journals (Sweden)

    Aydan ÖRSÇELİK

    2016-09-01

    Full Text Available Ankle sprain is one of the most common sports injuries. Chronic ankle instability is a common complication of ankle sprains. Two causes of chronic ankle instability are mechanical instability and functional instability. It is important to understand functional instability etiopathogenesis of the ankle joint in order to guide diagnosis and treatment. This article aims to understand the etiopathogenesis of functional ankle instability.

  12. Crack growth rate in core shroud horizontal welds using two models for a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis Juárez, C.R., E-mail: carlos.arganis@inin.gob.mx; Hernández Callejas, R.; Medina Almazán, A.L.

    2015-05-15

    Highlights: • Two models were used to predict SCC growth rate in a core shroud of a BWR. • A weld residual stress distribution with 30% stress relaxation by neutron was used. • Agreement is shown between the measurements of SCC growth rate and the predictions. • Slip–oxidation model is better at low fluences and empirical model at high fluences. - Abstract: An empirical crack growth rate correlation model and a predictive model based on the slip–oxidation mechanism for Stress Corrosion Cracking (SCC) were used to calculate the crack growth rate in a BWR core shroud. In this study, the crack growth rate was calculated by accounting for the environmental factors related to aqueous environment, neutron irradiation to high fluence and the complex residual stress conditions resulting from welding. In estimating the SCC behavior the crack growth measurements data from a Boiling Water Reactor (BWR) plant are referred to, and the stress intensity factor vs crack depth throughout thickness is calculated using a generic weld residual stress distribution for a core shroud, with a 30% stress relaxation induced by neutron irradiation. Quantitative agreement is shown between the measurements of SCC growth rate and the predictions of the slip–oxidation mechanism model for relatively low fluences (5 × 10{sup 24} n/m{sup 2}), and the empirical model predicted better the SCC growth rate than the slip–oxidation model for high fluences (>1 × 10{sup 25} n/m{sup 2}). The relevance of the models predictions for SCC growth rate behavior depends on knowing the model parameters.

  13. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

    International Nuclear Information System (INIS)

    Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

    2012-01-01

    The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

  14. A computerized operator support system for BWR power plant during normal and abnormal conditions

    International Nuclear Information System (INIS)

    Monta, K.; Sato, N.; Tsunoyama, S.; Sekimizu, K.; Mori, N.

    1984-01-01

    A computerized operator support system for BWR was developed based on the operators tasks which are divided into problem solving and controlling. Problem solving comprises fault managing and planning tasks. One of the operator supports for the problem solving is the automatic diagnostic support. However, the operator supports for controlling tasks are still important, e.g. to support its rule memory and rule execution. In the system presented here, operator supports are considered for both of these two categories. In the paper two subsystems for the latter are mainly described. (author)

  15. Proceedings of the international specialist meeting on BWR-pressure suppression containment technology. Vol. 1

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1981-01-01

    In the frame of R + D-work for BWR-pressure suppression systems the GKSS-Forschungszentrum Geesthacht GmbH organized an international specialist meeting. All important safety relevant aspects of pressure suppression system technology have been included. About 60 experts from USA, Japan, Sweden, Italy, Netherlands and the Federal Republic of Germany participated. They came from licensing authorities, vendors, research centers and universities. In 24 papers they have shown the world-wide present status of theoretical and experimental know-how on pressure suppression system behaviour. In discussions and working groups recommendations for future work have been compiled. (orig.) [de

  16. Connection between end plates and rods in a BWR fuel element

    International Nuclear Information System (INIS)

    Cali', G.P.

    1975-01-01

    The problem of the connection between the end plates and the rods of a BWR fuel element is analytically formulated. The behaviour of the springs coupling the rods with the upper plate is analyzed with particular detail since the deformation of these springs affects the forces at the interface of the fuel element structure components. A tool is given to design the springs according to some considerations regarding the mechanical strength of the interacting components as well as the influence of the possible geometrical unevennes of the system that can arise during the fuel element lifetime. (Cali', G.P.)

  17. The design and use of proficiency based BWR reactor maintenance and refuelling training mockups

    International Nuclear Information System (INIS)

    Ford, G.J.

    1996-01-01

    The purpose of this paper is to describe the ABB experience with the design and use of boiling water reactor training facilities. The training programs were developed and implemented in cooperation with the nuclear utilities. ABB operates two facilities, the ABB ATOM Light Water Reactor Service Center located in Vasteras, Sweden, and the ABB Combustion Engineering Nuclear Operations BWR Training Center located in Chattanooga, Tennessee, USA. The focus of the training centers are reactor maintenance and refueling activities plus the capability to develop and qualify tools, procedures and repair techniques

  18. Analysis of the integrity of the pressure vessel of the BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Silva Luna, O.

    1982-01-01

    The presssure vessel of a BWR type reactor was monitored for cracking during alternating events of its in-service life. The monitoring was to determine criticality of fractures catastrophic fractures and the velocity of fracture propagation. Detected cracks were evaluated as specified in ASME code section XI, of a minimum wall thickness of 2.5% crack growths were compared a) of 1/10 of the critical maximum size and b) at in-service inspection intervals according to ASME recommendations to be established at the Laguna Verde nuclear plant. Finally conclusions are made and discussed. (author)

  19. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments.

  20. LOFT test support branch data abstract report: one-sixth scale model BWR jet pump test

    International Nuclear Information System (INIS)

    Crapo, H.S.

    1979-01-01

    Pump performance data are presented for a 1/6 scale model jet pump in tests conducted at the LOFT Test Support Blowdown Facility. Steady-state subcooled pump characterization tests were performed over a wide range of forward and reverse flow conditions, both at room temperature, and at elevated temperature (555 0 K). Blowdown tests were also performed to obtain two-phase performance data in configurations simulating the flow patterns in the intact and broken loops of a BWR during a recirculation line break transient

  1. AREVA 10x10 BWR fuel experience feedback and on going upgrading

    International Nuclear Information System (INIS)

    Lippert, Hans Joachim; Rentmeister, Thomas; Garner, Norman; Tandy, Jay; Mollard, Pierre

    2008-01-01

    Established with engineering and manufacturing operations in the US and Europe, AREVA NP has been and is supplying nuclear fuel assemblies and associated core components to boiling water reactors worldwide, representing today more than 63 000 fuel assemblies. The evolution of BWR fuel rod arrays from early 6x6 designs to the 10x10 designs first introduced in the mid 1990's yielded significant improvements in thermal mechanical operating limits, critical power level, cold shutdown margin, discharge burnup, as well as other key operational capabilities. Since first delivered in 1992, ATRIUM T M 1 0 fuel assemblies have now been supplied to a total of 32 BWR plants in the US, Europe, and Asia resulting in an operating experience over 20 000 fuel assemblies. This article presents in detail the operational experience consolidated by these more than 20 000 ATRIUM T M 1 0 BWR assemblies already supplied to utilities. Within the different 10x10 fuel assemblies available, the Fuel Assembly design is chosen and tailored to the operating strategies of each reactor. Among them, the latest versions of ATRIUM T M a re ATRIUM T M 1 0XP and ATRIUM T M 1 0XM fuel assemblies which have been delivered to several utilities worldwide. The article details key aspects of ATRIUM T M 1 0 fuel assemblies in terms of reliability and performance. Special attention is paid to key proven features, ULTRAFLOW T M s pacer grids, the use of part length fuel rods (PLFRs) and their geometrical optimization, water channel and load chain, upgraded features available for inclusion with most advanced designs. Regular upgrading of the product has been made possible thanks to a continuous improvement process with the aim of further upgrading BWR fuel assembly performance and reliability. Regarding thermal mechanical behavior of fuel rods, chromia (Cr2O3) doped fuel pellets, described in Reference 1, well illustrate this improvement strategy to reduce fission gas release, increase power thresholds for PCI

  2. Analysis of results of AZTRAN and AZKIND codes for a BWR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L.; Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M.

    2016-09-01

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

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

    International Nuclear Information System (INIS)

    Christensen, P. la Cour.

    1979-08-01

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

  4. Urethral morphology and bladder instability.

    Science.gov (United States)

    Hausegger, K A; Fotter, R; Sorantin, E; Schmidt, P

    1991-01-01

    In order to calculate the relationship between Spinning top urethra (STU) and bladder instability 160 voiding cystourethrogramms (VCU), performed in 102 girls, have been analysed retrospectively. 28 girls had STU, 16 of those had bladder instability as well (57%). We could not find the highly positive correlation between unstable bladder and STU as reported by other authors, although there was a statistically positive correlation between STU and bladder instability. However the confidence interval was very broad (38%-75%). We conclude that bladder instability may contribute to STU but cannot render the etiological explanation for all cases. STU seems to be a polyetiological sign. In our opinion only the combination of STU and bladder instability has a diagnostic impact, since several therapeutical concepts are available in cases of bladder instability. The finding of STU in the VCU should alert the examiner's attention to functional disorders of the lower urinary tract. If no instability can be found, STU should be considered to be a normal variant.

  5. Urethral morphology and bladder instability

    Energy Technology Data Exchange (ETDEWEB)

    Hausegger, K.A.; Fotter, R.; Sorantin, E. (Graz Univ. (Austria). Radiologische Klinik); Schmidt, P. (Rehabilitationszentrum, Schallerbach (Austria))

    1991-05-01

    In order to calculate the relationship between Spinning top urethra (STU) and bladder instability 160 voiding cystourethrogramms (VCU), performed in 102 girls, have been analysed retrospectively. 28 girls had STU, 16 of those had bladder instability as well (57%). We could not find the highly positive correlation between unstable bladder and STU as reported by other authors, although there was a statistically positive correlation between STU and bladder instability. However the confidence interval was very broad (38%-75%). We conclude that bladder instability may contribute to STU but cannot render the etiological explanation for all cases. STU seems to be a polyetiological sign. In our opinion only the combination of STU and bladder instability has a diagnostic impact, since several therapeutical concepts are available in cases of bladder instability. The finding of STU in the VCU should alert the examiner's attention to functional disorders of the lower urinary tract. If no instability can be found, STU should be considered to be a normal variant. (orig.).

  6. Secondary Instability of Second Modes in Hypersonic Boundary Layers

    Science.gov (United States)

    Li, Fei; Choudhari, Meelan M.; Chang, Chau-Lyan; White, Jeffery A.

    2012-01-01

    Second mode disturbances dominate the primary instability stage of transition in a number of hypersonic flow configurations. The highest amplification rates of second mode disturbances are usually associated with 2D (or axisymmetric) perturbations and, therefore, a likely scenario for the onset of the three-dimensionality required for laminar-turbulent transition corresponds to the parametric amplification of 3D secondary instabilities in the presence of 2D, finite amplitude second mode disturbances. The secondary instability of second mode disturbances is studied for selected canonical flow configurations. The basic state for the secondary instability analysis is obtained by tracking the linear and nonlinear evolution of 2D, second mode disturbances using nonlinear parabolized stability equations. Unlike in previous studies, the selection of primary disturbances used for the secondary instability analysis was based on their potential relevance to transition in a low disturbance environment and the effects of nonlinearity on the evolution of primary disturbances was accounted for. Strongly nonlinear effects related to the self-interaction of second mode disturbances lead to an upstream shift in the upper branch neutral location. Secondary instability computations confirm the previously known dominance of subharmonic modes at relatively small primary amplitudes. However, for the Purdue Mach 6 compression cone configuration, it was shown that a strong fundamental secondary instability can exist for a range of initial amplitudes of the most amplified second mode disturbance, indicating that the exclusive focus on subharmonic modes in the previous applications of secondary instability theory to second mode primary instability may not have been fully justified.

  7. Genetic instability in human tumors.

    Science.gov (United States)

    Raptis, Stavroula; Bapat, Bharati

    2006-01-01

    Genetic, or genomic, instability refers to a series of observed spontaneous genetic changes occurring at an accelerated rate in cell populations derived from the same ancestral precursor. This is far from a new finding, but is one that has increasingly gained more attention in the last decade due to its plausible role(s) in tumorigenesis. The majority of genetic alterations contributing to the malignant transformation are seen in growth regulatory genes, and in genes involved in cell cycle progression and arrest. Genomic instability may present itself through alterations in the length of short repeat stretches of coding and non-coding DNA, resulting in microsatellite instability. Tumors with such profiles are referred to as exhibiting a mutator phenotype, which is largely a consequence of inactivating mutations in DNA damage repair genes. Genomic instability may also, and most commonly, results from gross chromosomal changes, such as translocations or amplifications, which lead to chromosomal instability. Telomere length and telomerase activity, important in maintaining chromosomal structure and in regulating a normal cell's lifespan, have been shown to have a function in both suppressing and facilitating malignant transformation. In addition to such direct sequence and structural changes, gene silencing through the hypermethylation of promoter regions, or increased gene expression through the hypomethylation of such regions, together, form an alternative, epigenetic mechanism leading to instability. Emerging evidence also suggests that dietary and environmental agents can further modulate the contribution of genetic instability to tumorigenesis. Currently, there is still much debate over the distinct classes of genomic instability and their specific roles in the initiation of tumor formation, as well as in the progressive transition to a cancerous state. This review examines the various molecular mechanisms that result in this genomic instability and the potential

  8. Rayleigh-Taylor instability of cylindrical jets with radial motion

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiang M. [GE Nuclear, Wilmington, NC (United States); Schrock, V.E.; Peterson, P.F. [Univ. of California, Berkeley, CA (United States)

    1995-09-01

    Rayleigh-Taylor instability of an interface between fluids with different densities subjected to accelleration normal to itself has interested researchers for almost a century. The classic analyses of a flat interface by Rayleigh and Taylor have shown that this type of instability depends on the direction of acceleration and the density differences of the two fluids. Plesset later analyzed the stability of a spherically symmetric flows (and a spherical interface) and concluded that the instability also depends on the velocity of the interface as well as the direction and magnitude of radial acceleration. The instability induced by radial motion in cylindrical systems seems to have been neglected by previous researchers. This paper analyzes the Rayleigh-Taylor type of the spherical case, the radial velocity also plays an important role. As an application, the example of a liquid jet surface in an Inertial Confinement Fusion (ICF) reactor design is analyzed.

  9. Determination Of Slope Instability Using Spatially Integrated Mapping Framework

    Science.gov (United States)

    Baharuddin, I. N. Z.; Omar, R. C.; Roslan, R.; Khalid, N. H. N.; Hanifah, M. I. M.

    2016-11-01

    The determination and identification of slope instability are often rely on data obtained from in-situ soil investigation work where it involves the logistic of machineries and manpower, thus these aspects may increase the cost especially for remote locations. Therefore a method, which is able to identify possible slope instability without frequent ground walkabout survey, is needed. This paper presents the method used in prediction of slope instability using spatial integrated mapping framework which applicable for remote areas such as tropical forest and natural hilly terrain. Spatial data such as geology, topography, land use map, slope angle and elevation were used in regional analysis during desktop study. Through this framework, the occurrence of slope instability was able to be identified and was validate using a confirmatory site- specific analysis.

  10. Preliminary model for core/concrete interactions. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Murfin, W.B.

    1977-08-01

    A preliminary model is described for computing the rate of penetration of concrete by a molten LWR core. Among the phenomena included are convective stirring of the melt by evolved gases, admixture of concrete decomposition products to the melt, chemical reactions, radiative heat loss, and variation of heat transfer coefficients with local pressure. The model is most applicable to a two-phase melt (metallic plus oxidic) having a fairly high metallic content.

  11. Recent control and instrumentation systems for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Fujii, Hiroaki; Higashikawa, Yuichi; Sato, Hideyuki

    1990-01-01

    For the needs of the more stable operation of nuclear power stations, the upgrading of the measurement and control system for BWRs has been promoted by positively introducing remarkably advancing electronic technology. Further, it is aimed at to construct the synthetic digitized measurement and control system for nuclear power stations to heighten the operation reliability in ABWRs. As the first step of the development in the synthetic digitization, the monitoring and control system for radioactive waste treatment was put in practical use for No.5 plant of Kashiwazaki, Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. Hitachi Ltd. has promoted the development and the application to actual plants of the measurement and control system for BWRs, in which digital control technology, optical information transmission technology and the operation-supporting technology using a computer were utilized. Hereafter, it is intended to expand the application of digital measurement and control aiming at improving the reliability, operation performance and maintainability. The nuclear power plant control complex with advanced man-machine interface-90 (NUCAMM-90) was developed, and its application to actual plants is planned. (K.I.)

  12. Sheared Electroconvective Instability

    Science.gov (United States)

    Kwak, Rhokyun; Pham, Van Sang; Lim, Kiang Meng; Han, Jongyoon

    2012-11-01

    Recently, ion concentration polarization (ICP) and related phenomena draw attention from physicists, due to its importance in understanding electrochemical systems. Researchers have been actively studying, but the complexity of this multiscale, multiphysics phenomenon has been limitation for gaining a detailed picture. Here, we consider electroconvective(EC) instability initiated by ICP under pressure-driven flow, a scenario often found in electrochemical desalinations. Combining scaling analysis, experiment, and numerical modeling, we reveal unique behaviors of sheared EC: unidirectional vortex structures, its size selection and vortex propagation. Selected by balancing the external pressure gradient and the electric body force, which generates Hagen-Poiseuille(HP) flow and vortical EC, the dimensionless EC thickness scales as (φ2 /UHP)1/3. The pressure-driven flow(or shear) suppresses unfavorably-directed vortices, and simultaneously pushes favorably-directed vortices with constant speed, which is linearly proportional to the total shear of HP flow. This is the first systematic characterization of sheared EC, which has significant implications on the optimization of electrodialysis and other electrochemical systems.

  13. Application of Galerkin's method for calculating boiling water reactor limit-cycle amplitude using the LAPUR feedback-transfer function and the point-kinetics equations

    International Nuclear Information System (INIS)

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

    1990-01-01

    This paper describes a technique for calculating boiling water reactor (BWR) behavior during steady-state limit-cycle oscillations. An approximate solution is obtained from the application of Galerkin's method to a BWR dynamic model consisting of the point-kinetics equations and the LAPUR-calculated power-to-reactivity feedback-transfer function. The approximate-solution technique is described, and comparisons of approximate solutions with numerical results and measured data are given. 7 refs., 5 figs

  14. 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.

  15. Atlantoaxial instability in Down's syndrome.

    OpenAIRE

    Collacott, R A

    1987-01-01

    The radiographs and clinical evaluations of 90 children with Down’s syndrome were reassessed after an interval of 5 years in a study of atlantoaxial instability (AAI) at the Derbyshire Children’s Hospital and Infirmary, Derby, UK.

  16. Summary of longitudinal instabilities workshop

    Energy Technology Data Exchange (ETDEWEB)

    Chasman, R.

    1976-01-01

    A five-day ISABELLE workshop on longitudinal instabilities was held at Brookhaven, August 9-13, 1976. About a dozen outside accelerator experts, both from Europe and the U.S.A., joined the local staff for discussions of longitudinal instabilities in ISABELLE. An agenda of talks was scheduled for the first day of the workshop. Later during the week, a presentation was given on the subject ''A more rigorous treatment of Landau damping in longitudinal beam instabilities''. A few progress meetings were held in which disagreements regarding calculations of coupling impedances were clarified. A summary session was held on the last day. Heavy emphasis was put on single bunched beam instabilities in the microwave region extending above the cut-off frequency of the ISABELLE vacuum chamber.

  17. Predicting Catastrophic BGP Routing Instabilities

    National Research Council Canada - National Science Library

    Nguyen, Lien

    2004-01-01

    .... Currently, this critical function is performed by the Border Gateway Protocol (BGP) version 4 RF01771. Like all routing protocols, BGP is vulnerable to instabilities that reduce its effectiveness...

  18. Genomic instability and radiation effects

    International Nuclear Information System (INIS)

    Christian Streffer

    2007-01-01

    Complete text of publication follows. Cancer, genetic mutations and developmental abnormalities are apparently associated with an increased genomic instability. Such phenomena have been frequently shown in human cancer cells in vitro and in situ. It is also well-known that individuals with a genetic predisposition for cancer proneness, such as ataxia telangiectesia, Fanconi anaemia etc. demonstrate a general high genomic instability e.g. in peripheral lymphocytes before a cancer has developed. Analogous data have been found in mice which develop a specific congenital malformation which has a genetic background. Under these aspects it is of high interest that ionising radiation can increase the genomic instability of mammalian cells after exposures in vitro an in vivo. This phenomenon is expressed 20 to 40 cell cycles after the exposure e.g. by de novo chromosomal aberrations. Such effects have been observed with high and low LET radiation, high LET radiation is more efficient. With low LET radiation a good dose response is observed in the dose range 0.2 to 2.0 Gy, Recently it has been reported that senescence and genomic instability was induced in human fibroblasts after 1 mGy carbon ions (1 in 18 cells are hit), apparently bystander effects also occurred under these conditions. The instability has been shown with DNA damage, chromosomal aberrations, gene mutation and cell death. It is also transferred to the next generation of mice with respect to gene mutations, chromosomal aberrations and congenital malformations. Several mechanisms have been discussed. The involvement of telomeres has gained interest. Genomic instability seems to be induced by a general lesion to the whole genome. The transmission of one chromosome from an irradiated cell to an non-irradiated cell leads to genomic instability in the untreated cells. Genomic instability increases mutation rates in the affected cells in general. As radiation late effects (cancer, gene mutations and congenital

  19. New innovative electrocoagulation (EC) treatment technology for BWR colloidal iron utilizing the seeding and filtration electronically (SAFETTM) system

    International Nuclear Information System (INIS)

    Denton, Mark S.; Bostick, William D.

    2007-01-01

    is 1) to break the colloid (i.e., break the outer radius repulsive charges of the similar charged colloidal particles), 2) allow these particles to now flocculate (floc), and 3) form a type of floc that is more readily filterable, and, thus, de-waterable. This task has been carried out with the innovative application of electronically seeding the feed stream with the metal of choice, and without the addition of chemicals common to ferri-flocking, or polymer addition. This patent-pending new system and technique is called Seeding And Filtration Electronically, or the SAFE TM System. Once the colloid has been broken and flocking has begun, removal of the resultant floc can be carried out by standard, back-washable (or, in simple cases, dead-end) filters; or simply in de-waterable HICs or liners. Such applications include low level radwaste (LLW) from both PWRs and BWRs, fuel pools, storage basins, salt water collection tanks, etc. For the removal of magnetic materials, such as some BWR irons, an Electro Magnetic Filter (EMF) was developed to couple with the Electro Coagulation (EC), (or metal-Flocking) Unit. In the advent that the waste stream primarily contains magnetic materials (e.g., boiler condensates and magnetite, and he-magnetite from BWRs), the material was simply filtered using the EMF. Bench-, pilot- and full-scale systems have been assembled and applied on actual plant waste samples quite successfully. The effects of initial feed pH and conductivity, as well as flocculation retention times was examined prior to applying the production equipment into the field. Since the initial studies (Denton, et al, EPRI, 2006), the ultimate success of field applications is now being demonstrated as the next development phase. For such portable field demonstrations and demand systems, a fully self enclosed (secondary containment) EC system was first developed and assembled in a modified B 25 Box (Floc-In-A-Box) and is being deployed to a number of NPP sites. Finally, a

  20. Microsatellite instability in bladder cancer

    DEFF Research Database (Denmark)

    Gonzalez-Zulueta, M; Ruppert, J M; Tokino, K

    1993-01-01

    Somatic instability at microsatellite repeats was detected in 6 of 200 transitional cell carcinomas of the bladder. Instabilities were apparent as changes in (GT)n repeat lengths on human chromosome 9 for four tumors and as alterations in a (CAG)n repeat in the androgen receptor gene on the X...... or larger (> 2 base pairs) alterations in repeat length. All six tumors were low stage (Ta-T1), suggesting that these alterations can occur early in bladder tumorigenesis....